• Journal of Radiation Industry
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• v.18 no.2
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• pp.133-140
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• 2024

The Korean 2^nd basic plan for management of high-level radioactive waste presented a plan to manage spent nuclear fuel through dry storage facilities in NPP on-site. For the construction and operation of the facility, it is necessary to develop the monitoring system of the integrity of spent nuclear fuel before operation. NUREG-1536 recommends that the theoretical cask array, typically in the 2×10 array, should account for shadowing effect among the dry storage casks. The objective of this study was to evaluate neutron flux accounting for shadowing effect among dry storage casks. The neutron release rate was evaluated using ORIGEN based on the design basis fuel condition. And the simulation of dry storage casks and evaluation of the shadowing effect were performed using MCNP. Shadowing effect of other dry storage casks was the highest at the center of the dry storage facility of the 2×10 array compared with the outside of the cask. The shadowing effect of neutron flux on the surface among the metal casks was approximately 18% at point 1, 23% at point 2, and 43% at point 3. For the concrete casks, the shadowing effect of neutron flux on the surface was approximately 46% at point 1, 51% at point 2, and 52% at point 3. This means that correction is necessary to monitor the integrity of spent nuclear fuel in each dry storage cask through evaluation of shadowing effect. The results of this study will be used for comparative analysis of neutron measurement data from spent nuclear fuels in dry storage cask. Additionally, the neutron flux evaluation procedure used in this study could be used as the basic data of safety assessment of dry storage cask and development of safety guide.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2356-2365
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• 2023

Characterization, dismantling and pre-disposal management of irradiated graphite (i-graphite) have an important role in safe decommissioning of several nuclear facilities which used this material as moderator and reflector. In addition to common radiation protection issues, easily volatizing long-lived radionuclides and stored Wigner energy could be released during imprudent retrieval and processing of i-graphite. With this regard, among all cutting technologies, abrasive waterjet (AWJ) can successfully achieve all of the thermo-mechanical and radiation protection objectives. In this work, factorial experiments were designed and systematically conducted to characterize the AWJ processing parameters and the machining capability. Moreover, the limitation of dust production and secondary waste generation has been addressed since they are important aspects for radiation protection and radioactive waste management. The promising results obtained on non-irradiated nuclear graphite blocks demonstrate the applicability of AWJ as a valid technology for optimizing the retrieval, storage, and disposal of such radioactive waste. These activities would benefit from the points of view of safety, management, and costs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.31-39
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• 2013

Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

• Journal of Korean Society of Environmental Engineers
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• v.37 no.5
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• pp.312-323
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• 2015

This paper presents future directions for spent nuclear fuel and high-level radioactive waste management. The successes and failures of siting nuclear waste repository experienced by the United States and other countries are reviewed with the current policy stance. Further, the needs for establishing management policy, considering the high-level radioactive waste produced by the dismantlement, nuclear security concerns, and cost-effectiveness analysis for the total nuclear fuel cycle, are emphasised. Technical discussions are organised into three main topics: interim storage, permanent disposal, and reprocessing. Licensing regimes are also investigated to suggest strategic plans for research and development programmes in the Republic of Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.245-253
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• 2006

A test of the 1/3 scale model was conducted to verify the tip-over analysis of a dry. concrete storage cask under a hypothetical accident condition. The tip-over analysis was executed using the velocity at each point as the initial conditions of the model just before the impact. The initial velocity was determined from the initial angular velocity, which would make the equivalent kinetic energy to the potential energy. To confirm the structural integrity of the canister, the visual testing and the non-detective testings such as Liquid Penetrant testing and Ultrasonic Testing were conducted. The lid of a storage cask was plastically deformed near the impact point. The structural integrity of storage cask was maintained. To verify the tip-over analysis the strains and the accelerations acquired by the tip-over test were compared with those by the analyses. The results of the analysis were larger than the test results about two times.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.231-236
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• 2011

When the Fukushima nuclear power plant accident occurred in March of 2011, a hydrogen explosion in the reactor building at the 4th unit of Fukushima plants led to a big surprise because the full core of the unit 4 reactor had been moved and stored underwater at the spent nuclear fuel storage pool for periodic maintenance. It was because the possible criticality in the fuel storage pool by coolant loss may yield more severe situation than the similar accident happened inside the reactor vessel. Fortunately, it was assured to be evitable to an anxious situation by a look of water filled in the storage pool later. In the paper, the safety state of the spent fuel storage pool and rack structures of the domestic nuclear plants would be roughly reviewed and compared with the Fukushima plant case by engineering viewpoint of potential severe accidents.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.165-175
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• 2005

Measurement of spent resin activity was initiated in 2004 in order to develop the C-14 removal technology for safe disposal. As part of this program, spent resins were sampled and measured in the in-station resin storage tank 2 at Wolsong Nuclear Power Plant Unit 1. At the time of sampling, the resins had been in storage tank from 3 to 23 years. Total 72 resin samples were sampled, which were collected from both man-hole (68 samples) and test-hole (4 samples) in the in-station resin storage tank 2. They were separated into liquid, activated carbon, zeolite, and spent resin. The spent resins were oxidized with sample oxidizer and analyzed for C-14. Ten of collected mixed resin samples were separated by density into cation and anion resins using a sugar solution. The C-14 concentration in anion exchange resin was approximately 2 times higher than in the mixed resin. The average concentration of C-14 in the cation/anion mixed exchange resin was $460\;GBq/m^3$ from test-hole and $53.1\;GBq/m^3$ from man-hole. We have found that concentration of C-14 in the spent resin is about from 0.4 to $1,321\;GBq/m^3$. So it could be a problem, when dispose of at a repository, since there is a disposal limit of $222\;GBq/m^3$. This means we should develop the C-14 removal technology.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.3
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• pp.307-322
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• 2021

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.442-448
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• 2022

Intermediate-level radioactive waste (ILW) is not subject to legal approval for cave disposal in Korea. To solve this problem, a spent resin treatment device that separates ¹⁴C-containing resin from zeolite/activated carbon and desorbs ¹⁴C through a microwave device has been developed. In this study, we evaluated the radiological safety of the operators performing repair work in the event of a failure in such a device treating 1 ton of spent resin mixture per day. Based on the safety evaluation results, it is possible to formulate a design plan that can ensure the safety of workers while developing a commercialized device. When each component of the resin treatment device can be repaired from the outside, the maximum and minimum allowable repair times are calculated as 263.2 h and 27.7 h for the ¹⁴C-detached resin storage tank and zeolite/activated carbon storage tank, respectively. For at least 6 h per quarter, the worker's annual dose limit remains within 50 mSv/year; further, over 5 years, it remained within 100 mSv. At least 6 h of repair time per quarter is considered, under conservative conditions, to verify the radiological safety of the worker during repair work within that time.

• Journal of computational fluids engineering
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• v.21 no.2
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• pp.47-53
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• 2016

This study investigated natural convection flow behavior in airflow path designed in concrete dry storage cask to remove the decay heat from spent nuclear fuels. Using FLUENT 16.1 code, thermal analysis for natural convection was carried out for three dimensional, 1/4 symmetry model under the normal condition that inlet ducts are 100% open. The maximum temperatures on other components except the fuel regions were satisfied with allowable values suggested in nuclear regulation-1536. From velocity and temperature distributions along the flow direction, the flow behavior in horizontal duct of air inlet and outlet duct, annular flow-path and bent pipe was delineated in detail. Theses results will be used as the theoretical background for the composing of airflow path for the designing of passive heat removal system by understanding the flow phenomena in airflow path.