• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.306-312
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• 2018

There is an increasing need in the United States and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations or interim storage sites. Under normal conditions, the Nuclear Regulatory Commission limits cladding temperature to $400^{\circ}C$ for high-burnup (>45 GWd/mtU) fuel, with higher temperatures allowed for low-burnup fuel. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at $400^{\circ}C$. Results were representative of the majority of US light water reactor fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

• 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.45 no.3
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• pp.377-384
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• 2013

During the last three decades, South Korean nuclear power plants have discharged about 5,950 tons of spent fuel and the maximum burn-up reached 55 GWd/MTU in 2002. This study was performed to support the development of Korean dry spent fuel storage alternatives. First, we chose V5H-$17{\times}17$ and KSFA-$16{\times}16$ as representative domestic spent fuels, considering current accumulation and the future generation of the spent fuels. Examination reveals that their average burn-ups have already increased from 33 to 51 GWd/MTU and from 34.8 to 48.5 GWd/MTU, respectively. Evaluation of the fuel characteristics shows that at the average burn-up of 42 GWd/MTU, the oxide thickness, hydrogen content, and hoop stress ranged from $30{\sim}60{\mu}m$, 250 ~ 500 ppm, and 50 ~ 75 MPa, respectively. But when burn-up exceeds 55 GWd/MTU, those characteristics can increase up to 100 ${\mu}m$, 800 ppm, and 120 MPa, respectively, depending on the power history. These results demonstrate that most Korean spent nuclear fuels are expected to remain within safe bounds during long-term dry storage, however, the excessive hoop stress and hydrogen concentration may trigger the degradation of the spent fuel integrity early during the long-term dry storage in the case of high burn-up spent fuels exceeding 45 GWd/MTU.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.531-541
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• 2023

The Wolsong Nuclear Power Plant (NPP) operates an on-site spent fuel dry storage facility using concrete silo and vertical module systems. This facility must be safely maintained until the spent nuclear fuel (SNF) is transferred to an external interim or final disposal facility, aligning with national policies on spent nuclear fuel management. The concrete silo system, operational since 1992, requires an aging management review for its long-term operation and potential license renewal. This involves comparing aging management programs of different dry storage systems against the U.S. NRC's guidelines for license renewal of spent nuclear fuel dry storage facilities and the U.S. DOE's program for long-term storage. Based on this comparison, a specific aging management program for the silo system was developed. Furthermore, the facility's current practices-periodic checks of surface dose rate, contamination, weld integrity, leakage, surface and groundwater, cumulative dose, and concrete structure-were evaluated for their suitability in managing the silo system's aging. Based on this review, several improvements were proposed.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.614-620
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• 2020

In a previous study, delayed hydride cracking (DHC) assessment of pressurized water reactor (PWR) spent fuel during dry storage using the threshold stress intensity factor (K_IH) was performed. However, there were a few limitations in the analysis of the cladding properties, such as oxide thickness and mechanical properties. In this study, those models were modified to include test data for irradiated materials, and the cladding creep model was introduced to improve the reliability of the DHC assessment. In this study, DHC susceptibility of PWR spent fuel during dry storage depending on the axial elevation was evaluated with the improved assessment methodology. In addition, the sensitivity of affecting parameters such as fuel burnup, hydride thickness, and crack aspect ratio are presented.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.489-498
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• 2011

UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

• Journal of the Korea institute for structural maintenance and inspection
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• v.9 no.2
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• pp.95-102
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• 2005

Delaying and objection for the construction of storage spent-fuel disposal has prompted to consider expanding on-site storage of spent reactor fuel since it can eliminate the need for costly and difficult shipping and control of the spent fuel completely under the direction of the owner-utility. The dry storage unit developed in Canada can accommodate Korea heavy water reactor fuel elements and become a candidate for the Korean market. In this paper, finite element analysis were carried out in order to investigate the structural behavior of the nuclear spent fuel dry storage system, which is subjected to impact loads such as collision of a truck load and dropping of flask under the irregular operation.

• Journal of Welding and Joining
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• v.23 no.6
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• pp.43-48
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• 2005

This work was carried out to obtain sound welds and to select a most suitable binary metal joint among three different dissimilar binary metal combinations such as Zr-4/Ta, Mo/Ta and Ti/Ta(seal tube/sensor sheath) joints fur the instrumented nuclear fuel irradiation test. To do this, Taguchi experimental method was employed to optimize the experimental data. In addition, metallography, micro-focus x-ray radiography and hardness test were conducted to examine the welds. From the weld bead appearance, penetration depth and bead width as well as weld defects standpoint, Zr-4/Ta joint is suggested for the circumferential joining between a seal tube and a sensor sheath. The optimized welding parameters based on Zr-4/Ta joint are suggested as well.

• Proceedings of the Korea Concrete Institute Conference
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• 2004.11a
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• pp.381-384
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• 2004

Delaying and objection for the construction of storage spent-fuel disposal has prompted to consider expanding on-site storage of spent reactor fuel since it can eliminate the need for costly and difficult shipping and control of the spent fuel completely under the direction of the owner-utility. The dry storage unit developed in Canada can accommodate Korea heavy water reactor fuel elements and become a candidate for the Korean market. In this paper, finite element analyses were carried out in order to investigate the structural behavior of the nuclear spent fuel dry storage system, which is subjected to impact loads such as collision of a truck load and dropping of flask under the irregular operation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.2
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• pp.117-128
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• 2024

The 300 concrete silo systems installed and operated at the site of Wolsong nuclear power plant (NPP) have been storing CANDU spent nuclear fuel (SNF) under dry conditions since 1992. The dry storage system must be operated safely until SNF is delivered to an interim storage facility or final repository located outside the NPP in accordance with the SNF management policy of the country. The silo dry storage system consists of a concrete structure, liner steel plate in the inner cavity, and fuel basket. Because the components of the silo system are exposed to high energy radiation owing to the high radioactivity of SNF inside, the effects of irradiation during long-term storage must be analyzed. To this end, material specimens of each component were manufactured and subjected to irradiation and strength tests, and mechanical characteristics before and after irradiation were examined. Notably, the mechanical characteristics of the main components of the silo system were affected by irradiation during the storage of spent fuel. The test results will be used to evaluate the long-term behavior of silo systems in the future.