• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.709-716
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• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.609-613
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• 2003

The confinement of a storage system to accommodate spent fuel and radioactive material must be designed and estimated so that the storage system is safe during a storage period. The confinement can be estimated by calculation of the allowable leakage rate in compliance with ANSI Nl4.5 or ISO 12807. Accordingly, the allowable leakage rate was estimated in compliance with ISO 12807 in the case of storage of 24 PWR spent fuels and 24-conditioned spent fuels. In the case of the 24 PWR spent fuel assemblies, the allowable leakage rate was estimated as $7.43{\times}10_{-11}m_3/s$, in the case of the 24 conditioned spent fuel assemblies, the allowable leakage rate was estimated as $1.80{\times}10_{-10}m_3/s$. Therefore, the confinement condition in the storage of the conditioned spent fuel is easier than that for the storage of the PWR spent fuel.

• 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.17 no.1
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• pp.1-7
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• 1985

Non-destructive gamma-ray spectrometry was carried out on the spent PWR fuel assemblies at the spent fuel pool of reactor-site. Attention was focused on the determination of burnup distribution and cooling time. For the measurement of burnup distribution, the concentration ratio of $^{134}$ Cs$^{137}$ Cs was used and the results showed these ratios varied with the positions of assemblies in the core during their irradiation. For the measurement of cooling time, $^{144}$ Ce$^{137}$ Cs was used and the results were agreed considerably well with the operator declared cooling time.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.504-511
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• 2018

In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for "Channelized Large Power Reactor"-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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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.

• 제어로봇시스템학회:학술대회논문집
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• 1987.10b
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• pp.613-618
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• 1987

Rod consolidation is a method of increasing spent nuclear fuel storage capacity by disassembling fuel assemblies thus storing the fuel rods in a tighter array. It involves some basic operations which closely resemble to the material handling of a manufacturing process. But all the operations must be controlled remotely in shielded environment from outside due to the highly radioactive nature of the workpiece. In this study the status of the rod consolidation technology in other countries has been surveyed and a feasibility study for the conceptual design of this process have been made.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.601-607
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• 2022

In recent years, monitoring of spent nuclear fuel inside dry cask storage has become an important area of national security. Muon tomography is a useful method for monitoring spent nuclear fuel because it uses high energy muons that penetrate deep into the target material and provides a 3-D structure of the inner materials. We designed a muon tomography system consisting of four 2-D position sensitive detector and characterized and optimized the system parameters. Each detector, measuring 200 × 200 cm², consists of a plastic scintillator, wavelength shifting (WLS) fibers and, SiPMs. The reconstructed image is obtained by extracting the intersection of the incoming and outgoing muon tracks using a Point-of-Closest-Approach (PoCA) algorithm. The Geant4 simulation was used to evaluate the performance of the muon tomography system and to optimize the design parameters including the pixel size of the muon detector, the field of view (FOV), and the distance between detectors. Based on the optimized design parameters, the spent fuel assemblies were modeled and the line profile was analyzed to conduct a feasibility study. Line profile analysis confirmed that muon tomography system can monitor nuclear spent fuel in dry storage container.