• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.17-26
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• 1986

A test is made for a method to determine reliable bias in the criticality safety analysis of spent fuel storage pool with turnup credit between the reference and rack criticality calculation methods. The spent fuel pool of Kori Unit 1 is conceptually redesigned to the most compact rack with turnup credit, and its multiplication factors are calculated depending on fuel enrichment and burnup, by the Monte Carlo code KENO-IV as a reference and by a two-dimensional collision probability code FATAC as a practical method. Then, the computed values with the help of the above two computer codes are compared to evaluate the bias and its trend in terms of multiplication factor on fuel enrichment and turnup. The result indicates that the bias can be determined with reliability basis but without any disadvantage in criticality safety margin compared with the conventional method.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.43-49
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• 2022

The purpose of this study is to evaluate the radiation effect on damage when the external shield of the spent nuclear fuel transport cask is damaged due to impact as the cause of an unexpected accident. The neutron and gamma-ray intensities and spectra are calculated using the ORIGEN-Arp module in the SCALE 6.2.4 code package⁽¹⁾ and then using MCNP6.2⁽²⁾ code calculate the dose rate. In order to evaluate the radiation dose according to the size of damage caused by external impact, various sized holes of 0.3~13.7% are assumed in the outer shield of the cask to evaluate the sensitivity to the dose. In the case of radiation source leakage, damage to the nuclear fuel assembly is assumed to be up to 6% based on overseas test cases. When only the outer shield is damaged, the maximum surface dose is calculated as 3.12E+03 mSv/hr. However, if the radiation source is leaked due to damage to the nuclear fuel assembly, it becomes 7.00E+05 mSv/hr which is about 200 times greater than the former case.

• The Journal of the Korea institute of electronic communication sciences
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• v.14 no.3
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• pp.505-512
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• 2019

We propose a new method to monitor the water level of spent fuel pool in a nuclear power plant without electric power. We also analyze the performance and limitation of the proposed method. Our method is based on the reflected optical power at the end of optical fiber through a $1{\times}N$ optical coupler. We reveal that there is no problem to monitor the water level when using a $1{\times}8$ optical coupler. However, when a $1{\times}16$ optical coupler is used, only 15 out of 16 output ports can be used due to Rayleigh back-scattering. When a $1{\times}32$ optical coupler is used, only 25 out of 32 output ports can be used to monitor the water level.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.237-245
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• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.788-793
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• 2018

High-density spent fuel (SF) storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.627-638
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• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.459-470
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• 2012

From the PSA point of view, the Fukushima accident of Japan in 2011 reveals some issues to be re-considered and/or improved in the PSA such as the limited scope of the PSA, site risk, etc. KAERI (Korea Atomic Energy Research Institute) has performed researches on the development of an integrated risk assessment framework related to some issues arisen after the Fukushima accident. This framework can cover the internal PSA model and external PSA models (fire, flooding, and seismic PSA models) in the full power and the low power-shutdown modes. This framework also integrates level 1, 2 and 3 PSA to quantify the risk of nuclear facilities more efficiently and consistently. We expect that this framework will be helpful to resolve the issue regarding the limited scope of PSA and to reduce some inconsistencies that might exist between (1) the internal and external PSA, and (2) full power mode PSA and low power-shutdown PSA models. In addition, KAERI is starting researches related to the extreme external events, the risk assessment of spent fuel pool, and the site risk. These emerging issues will be incorporated into the integrated risk assessment framework. In this paper the integrated risk assessment framework and the research activities on the emerging issues are outlined.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.435-443
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• 2016

This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.411-422
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• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.91-98
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• 1998

Time history analysis is usually performed to characterize the nonlinear seismic behavior of a spent fuel storage rack(SFSR). In the past, the seismic analyses of the SFSR were performed with two-dimensional planar models, which could not account for torsional response and simultaneous multi-directional seismic input In this study, three-dimensional seismic analysis methodology is developed for the single SFSR using the ANSYS code. The 3D- Model can be used to determine the nonlinear behavior of the rack, i.e., sliding, uplifting, and impact evaluation between the fuel assembly and rack, and rack and the pool wall, This paper also reviews the 3-D modeling of the SFSR and the adequacy of the ANSYS for the seismic analysis. AS a result of the adquacy study, the method of ANSYS transient analysis with acceleration time history is suitable for the seismic analysis of highly nonlinear structure such as an SFSR but it isn't appropriate to use displacement time history of seismic input.