• Nuclear Engineering and Technology
• /
• v.45 no.4
• /
• pp.427-438
• /
• 2013

In this paper we discuss spent fuel management options in the Republic of Korea (ROK) from two interrelated perspectives: Centralized dry cask storage and spent fuel pyroprocessing and burning in sodium fast reactors (SFRs). We argue that the ROK will run out of space for at-reactors spent fuel storage by about the year 2030 and will thus need to transition centralized dry cask storage. Pyroprocessing plant capacity, even if approved and successfully licensed and constructed by that time, will not suffice to handle all the spent fuel discharged annually. Hence centralized dry cask storage will be required even if the pyroprocessing option is successfully developed by 2030. Pyroprocessing is but an enabling technology on the path leading to fissile material recycling and burning in future SFRs. In this regard we discuss two SFR options under development in the U.S.: the Super Prism and the Travelling Wave Reactor (TWR). We note that the U.S. is further along in reactor development than the ROK. The ROK though has acquired more experience, recently in investigating fuel recycling options for SFRs. We thus call for two complementary joint R&D project to be conducted by U.S. and ROK scientists. One leading to the development of a demonstration centralized away-fromreactors spent fuel storage facility. The other involve further R&D on a combined SFR-fuel cycle complex based on the reactor and fuel cycle options discussed in the paper.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.3
• /
• pp.245-253
• /
• 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)
• /
• v.22 no.2
• /
• pp.201-210
• /
• 2024

Because most spent nuclear fuel storage casks have been designed for low burnup fuel, a safety-significant high burnup dry storage cask must be developed for nuclear facilities in Korea to store the increasing high burnup and damaged fuels. More than 20% of fuels generated by PWRs comprise high burnup fuels. This study conducted a structural safety evaluation of the preliminary designs for a high burnup storage cask with 21 spent nuclear fuels and evaluated feasible loading conditions under normal, off-normal, and accident conditions. Two types of metal and concrete storage casks were used in the evaluation. Structural integrity was assessed by comparing load combinations and stress intensity limits under each condition. Evaluation results showed that the storage cask had secured structural integrity as it satisfied the stress intensity limit under normal, off-normal, and accident conditions. These results can be used as baseline data for the detailed design of high burnup storage casks.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.2
• /
• pp.271-279
• /
• 2018

During the seven years from 2009 to 2016, PWR SNF (spent nuclear fuel) transportation and storage systems suitable for domestic conditions were developed by the government to cope with the saturation of wet storage capacity in NPPs. One of the developed systems is a multipurpose metal cask applicable for transportation/storage; the other is a concrete cask dedicated to storage. Efficient cask technologies were secured utilizing the characteristics and experience of relevant industrial, academic and research institutes. Technological independence was also achieved through several patent registrations of research outcomes. To prepare for a rapid increase of demand in the near future, technology transfer of secured patents and technologies to the domestic industry was carried out twice in the years of 2016 and 2017.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.4
• /
• pp.455-472
• /
• 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 Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.1
• /
• pp.13-22
• /
• 2004

This study presents the thermal analyses of a spent fuel dry storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 15 $^{\circ}C$ under the normal condition. The off-normal condition has an environmental temperature of 38 $^{\circ}C$. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Two of the four air inlet ducts are assumed to be completely blocked. The significant thermal design feature of the storage cask is the air flow path used to remove the decay heat from the spent fuel. Natural circulation of the air inside the cask allows the concrete and fuel cladding temperatures to be maintained below the allowable values. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. The maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal and off-normal conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.4
• /
• pp.435-443
• /
• 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)
• /
• v.4 no.2
• /
• pp.139-152
• /
• 2006

A spent fuel storage cask is required to prove the safety of a canister under a hypothetical accidental drop condition. A hypothetical accidental drop condition means that a canister is assumed to be a lee drop on to a pad of the storage cask during loading it into a storage cask. A pad of the storage cask absorbs shock to maintain the structural integrities of a canister under a hypothetical accidental drop condition. In this paper a finite element analysis for various pad structures was carried out to improve the structural integrity of a canister under a hypothetical accidental drop condition. A pad of a storage cask was designed a steel structure with concrete. The 1/4 height of a pad was modified with a structure composed of a steel and a polyurethane foam as a impact limiter. The effect of a shape of a steel structure was studied. The effects of the thickness of a steel structure and the density of a polyurethane foam was also studied.

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

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 2005.04a
• /
• pp.383-389
• /
• 2005

This research is to develop a seismic response analysis method for a spent fuel storage cask. FEM model is built for the test model of 1/8 scale spent fuel dry storage cask using available 3D contact conditions in ABAQUS/Explicit. Input load for this analysis os a seismic wave of El-centro earthquake, and the friction and damping coefficients in the analysis condition we obtained from the test result. Penalty and kinematic contact methods of ABAQUS are used for mechanical contact formulation. The analysis method was verified for rocking angle obtained by seismic response tests. The kinematic contact method with an adequate normal contact stiffness showed a good agreement with tests.