• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.179-186
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• 2009

This paper presents a thermal creep test under internal pressure and post-test characterization performed on high burnup (68 MWd/kgU) ZIRLO. This research has been done by the CSN, ENRESA, and ENUSA in order to investigate the behavior of advanced cladding materials in contemporary PWRs at higher burnup under dry cask storage conditions. Also, to investigate the hydride reorientation, the cool-down of the samples after the test has been done in a coordinated manner with the internal pressure. The creep results obtained are consistent with the expected behavior from reference CWSR material, Zr-4. During the test, the material retained significant ductility: one specimen leaked during the test at an engineering strain of the tube section of 17%; remarkably, the crack closed due to de-pressurization. Although significant hydride reorientation occurred during the cool-down under pressure, no specimen failed during the cool-down.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2013.05a
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• pp.155-156
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• 2013

본 연구는 사용후핵연료 수송 저장 용기인 저합금강(SA350 LF3)에 일정의 감마선을 조사하고 감마선 조사 전후 물성 및 내식 특성 변화와 표면처리에 의한 내식성 개선 효과에 관하여 연구하였다. 상온 항복강도 및 인장강도의 기계적 물성은 감마선 조사 여부에 따라 물성의 차이는 보이지 않았지만, 저온충격 특성은 감마선 조사를 하지 않은 충격 흡수에너지에 비교하여 조사후 시험편의 충격 흡수 에너지가 감소되었다. 양분극 곡선에 측정에 의해 관찰된 저합금강의 내식성은 감마선 조사된 시험편에서 감마선을 조사하지 않은 시험편 보다 낮은 부식전위를 나타내었다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.455-472
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• 2018

It is important to study how to manage dry storage casks of spent nuclear fuels (SNF), because wet storage spaces for SNF will shortly be at full capacity in the Republic of Korea. The US has operated a dry storage cask system for several decades, and has carried out significant studies into how to successfully manage dry storage cask for SNF. This type of expertise and experience is currently lacking in the Republic of Korea. The degradation of dry casks is an important issue that must be considered. In particular, chloride-induced stress corrosion cracking (CISCC) is known to lead to the release of radioisotopes from canisters. The U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the Electric Power Research Institute have undertaken research into the CISCC mechanism. In addition, Sandia National Laboratories (SNL) has extensively researched CISCC and how to manage it in dry storage canisters. In this review paper, the probabilistic model proposed by the SNL is analyzed and, based on this model, US-based CISCC research is reviewed in detail. This paper will inform the management of dry cask storage of SNF from light water reactors in austenite stainless steel canisters in the Republic of Korea.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.92-100
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• 2015

For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.12
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• pp.815-824
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• 2016

The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermal-hydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermal-hydraulic technology and the commercialization.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1009-1014
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• 2023

Muon tomography is a useful method for monitoring special nuclear materials (SNMs) such as spent nuclear fuel inside dry cask storage. Multiple Coulomb scattering of muons can be used to provide information about the 3-dimensional structure and atomic number(Z) of the inner materials. Tomography using muons is less affected by the shielding material and less harmful to health than other measurement methods. We developed a muon detector for muon tomography, which consists of a plastic scintillator, 64 long wavelength-shifting (WLS) fibers attached to the top of the plastic scintillator, and silicon photomultipliers (SiPMs) connected to both ends of each WLS fiber. The muon detector can acquire X and Y positions simultaneously using a position determination algorithm. The design parameters of the muon detector were optimized using DETECT2000 and Geant4 simulations, and a muon detector prototype was built based on the results. Spatial resolution measurement was performed using simulations and experiments to evaluate the feasibility of the muon detector. The experimental results were in good agreement with the simulation results. The muon detector has been confirmed for use in a muon tomography system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.237-245
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• 2020

Metal seals are the main components that establish the containment boundary in bolted casks, which store spent nuclear fuel. These seals are degraded by heat and radiation. In addition, creep occurs when the seals are exposed to intense heat for an extended period. This creep results in the stress relaxation of the seals, which consequently impairs the seal integrity. The stress relaxation can reduce the sealing performance of the metal seal, which can further cause leakage in the storage cask. Moreover, the reduction of bolt tension leads to sealing performance degradation. In this study, the results of high-temperature-accelerated tests were obtained to evaluate the containment integrity of metal seals and the decrease in bolt tension. During the tests, the leakage rate, bolt strain, and ambient temperature of the metal seals were measured and analyzed. The metal seals were found to maintain containment integrity for 50 years of storage. The validity of the acceleration test was also investigated.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3745-3765
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• 2022

The concrete structures related to nuclear safety are threatened by accidental impact loadings, mainly including the low-velocity drop-weight impact (e.g., spent fuel cask and assembly, etc. with the velocity less than 20 m/s) and high-speed projectile impact (e.g., steel pipe, valve, turbine bucket, etc. with the velocity higher than 20 m/s), while the existing studies are still limited in the impact resistant design of nuclear power plant (NPP), especially the primary RC slab. This paper aims to propose the numerical simulation and theoretical approaches to assist the impact-resistant design of RC slab in NPP. Firstly, the continuous surface cap (CSC) model parameters for concrete with the compressive strength of 20-70 MPa are fully calibrated and verified, and the refined numerical simulation approach is proposed. Secondly, the two-degree freedom (TDOF) model with considering the mutual effect of flexural and shear resistance of RC slab are developed. Furthermore, based on the low-velocity drop hammer tests and high-speed soft/hard projectile impact tests on RC slabs, the adopted numerical simulation and TDOF model approaches are fully validated by the flexural and punching shear damage, deflection, and impact force time-histories of RC slabs. Finally, as for the two low-velocity impact scenarios, the design procedure of RC slab based on TDOF model is validated and recommended. Meanwhile, as for the four actual high-speed impact scenarios, the impact-resistant design specification in Chinese code NB/T 20012-2019 is evaluated, the over conservation of which is found, and the proposed numerical approach is recommended. The present work could beneficially guide the impact-resistant design and safety assessment of NPPs against the accidental impact loadings.