• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.11-18
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• 2022

In this research, curium oxide (CmO₂) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO₂ is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO₂ and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO₂ is competent like UO₂ as a fuel due to higher fission cross-section of ²⁴³Cm and ²⁴⁵Cm isotopes and neutron capture cross-section of ²⁴⁴Cm and ²⁴⁶Cm isotopes. As a result, when some or all of the UO₂ of fuel rods or IBA rods are replaced by CmO₂, we get a similar k-inf like the reference even with lower enrichment UO₂ fuels. These studies show that the use of CmO₂ as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO₂ with inert material ZrO₂ in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2095-2103
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• 2021

A small power ADS design using thorium oxide and diluent matrix reprocessed fuel is proposed for a high transmutation rate, small reactivity swing, and strong safety features. Two fuel matrices (CERCER and CERMET) and different recycled fuel compositions recovered from UO₂ spent fuels with 45 GWd/tU and 60 GWd/tU burnup were investigated to determine the suitable fuel for the ADS. It was found that the transmutation of each isotope depends on TRU initial loading amount. After examining the cores, the results show that CERCER fueled ADS has a negative coolant void reactivity (CVR) and a smaller radiotoxicity at discharge compared to that of CERMET core. It implies that CERCER fuel has enhanced safety features and more flavor in terms of radiotoxicity management. To increase fuel utilization and core operation efficiency, a simple assembly shuffling pattern for the CERCER fueled ADS is also proposed. Eigenvalue and burnup calculations were conducted using Serpent 2 with ENDF/B-VII.0 library in both kcode and external source modes, and it indicates that the results of transmutation analyses obtained by kcode only is reliable to discuss the transmutation potential of ADS. Burnup calculation with the fixed-source mode is essential to be used for more practical results of the transmutation by ADS.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4647-4658
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• 2023

In this paper, elastic-plastic fracture mechanics analysis is performed to determine the critical crack sizes of the multipurpose canister (MPC) manufactured using austenitic stainless steel under dynamic loading conditions that simulate drop accidents. Firstly, dynamic finite element (FE) analysis is performed using Abaqus v.2018 with the KORAD (Korea Radioactive Waste Agency)-21 model under two drop accident conditions. Through the FE analysis, critical locations and through-thickness stress distributions in the MPC are identified, where the maximum plastic strain occurs during impact loadings. Then, the evaluation using the failure assessment diagram (FAD) is performed by postulating an external surface crack at the critical location to determine the critical crack depth. It is found that, for the drop cases considered in this paper, the principal failure mechanism for the circumferential surface crack is found to be the plastic collapse due to dominant high bending axial stress in the thickness. For axial cracks, the plastic collapse is also the dominant failure mechanism due to high membrane hoop stress, followed by the ductile tearing analysis. When incorporating the strain rate effect on yield strength and fracture toughness, the critical crack depth increases from 10 to 20%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.121-133
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• 2014

If the spent fuels or the high-level radioactive wastes can be disposed of in the depth of 3~5 km and more stable rock formation, it has several advantages. For example, (1)significant fluid flow through basement rock is prevented, in part, by low permeability, poorly connected transport pathways, and (2)overburden self-sealing. (3)Deep fluids also resist vertical movement because they are density stratified and reducing conditions will sharply limit solubility of most dose-critical radionuclides at the depth. Finally, (4) high ionic strengths of deep fluids will prevent colloidal transport. Therefore, as an alternative disposal concept to the deep geological disposal concept(DGD), very deep borehole disposal(DBD) technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, for the preliminary applicability analyses of the DBD system for the spent fuels or high level wastes, the DBD concepts which have been developed by some countries according to the rapid advance in the development of drilling technology were reviewed. To do this, the general concept of DBD system was checked and the study cases of foreign countries were described and analyzed. These results will be used as an input for the analyses of applicability for DBD in Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.75-87
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• 2020

The Korea Atomic Energy Research Institute (KAERI) has developed geological repository systems for the disposal of high-level wastes and spent nuclear fuels (SNFs) in South Korea. The purpose of the most recently developed system, the improved KAERI Reference Disposal System Plus (KRS⁺), is to dispose of all SNFs in Korea with improved disposal area efficiency. In this paper, a system-level safety assessment model for the KRS⁺ is presented with long-term assessment results. A system-level model is used to evaluate the overall performance of the disposal system rather than simulating a single component. Because a repository site in Korea has yet to be selected, a conceptual model is used to describe the proposed disposal system. Some uncertain parameters are incorporated into the model for the future site selection process. These parameters include options for a fractured pathway in a geosphere, parameters for radionuclide migration, and repository design dimensions. Two types of SNF, PULS7 from a pressurized water reactor and Canada Deuterium Uranium from a heavy water reactor, were selected as a reference inventory considering the future cumulative stock of SNFs in Korea. The highest peak radiological dose to a representative public was estimated to be 8.19×10^-4 mSv·yr^-1, primarily from ¹²⁹I. The proposed KRS⁺ design is expected to have a high safety margin that is on the order of two times lower than the dose limit criterion of 0.1 mSv·yr^-1.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.27-34
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• 1988

Since the lack of the spent fuel storage capcity has been expected for all Korean nuclear power plants in the mid-1990s, the maximum density rack (MDR) with consolidated fuels can be proposed to overcome the shortage of the storage capacity in KNU 9 & 10 which have most limited capacities. To ensure the safety when the alternatives are applied in the KNU 9 & 10, the multiplication factor are calculated with varying the rack pitch and the thickness of consolidated storage box by the AMPX-KENO IV codes. The computing system is verified by the benchmark calculation with criticality experiments for arrays of consolidated fuel modules, which was reported by B & W in 1981. Also an abnormal condition, i.e. malposition accident, is simulated. The results indicate that the KNU 9 & 10 storage pools with consolidated fuel are safe in the view of the criticality. Thus the storage capacity can be expanded from 9/3 cores into 27/3 cores even with considering equipments and cooling spaces.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.119-128
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• 2008

Thermal assessment of a new CANDU spent fuel disposal system, which improves the retrievability of the spent fuel and enhances the densification factor compared with the Korean Reference disposal System, is carried out in this study. The canisters for CANDU spent fuels are stored for long term and cooled by natural convection in the proposed disposal system for the retrievability. The steady state thermal analyses for proposed CANDU disposal system are carried out with the ANSYS 10.0 CFX code. The thermal analyses are performed through two steps. At the first step, the sensitivity of the disposal tunnel spacing is analysed. The differences of maximum temperatures by several tunnel spacings are calculated at three points in the disposal tunnel. The result shows that the differences of the temperature at the three points are almost negligible because 99% of the decay heat is removed by natural convection. At the second procedure, 60m tunnel spacing with a ventilation system instead of natural convection is considered. The result is applied to the calculation of the canister surface temperature in disposal tunnel as boundary conditions. Consequently, the average and the maximum surface temperature of disposal canisters are $79.9^{\circ}C$ and $119^{\circ}C$, respectively. The inner maximum temperature of a basket in the disposal canister is calculated as $140.9^{\circ}C$. The maximum temperature of the basket meets the thermal requirement for the CANDU spent fuel cladding.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.289-295
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• 2016

Kori Unit 1, the first PWR (Pressurized Water Reactor) plant in Korea, began its commercial operation in 1978 and will permanently shut down on June 18, 2017. After moving the spent fuels to SFP (Spent Fuel Pool) system, Kori Unit 1 will perform a full system chemical decontamination to reduce radiation levels inside the various plant systems. This paper will describe the operation concept of the full system chemical decontamination for Kori Unit 1 based on experiences overseas.

• Analytical Science and Technology
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• v.13 no.6
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• pp.751-758
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• 2000

A study was carried out on the dissolution of spent PWR fuels and performed on the fuels and the separation of iodide for the quantitative analysis using SIMFUEL which has chemical composition of a simulated spent PWR fuel (burn-up; 35,000 MWd/MTU and cooling time; 10 years). To dissolve the SIMFUEL effectively and to minimize the formation of volatile iodine through dissolution process, the optimum ratio of mixed acid ($HNO_3/HCl$ 80: 20 mol%) was established and ozone gas was purged. In the separation step of iodine with $CCl_4$, $NH_2OH{\cdot}HCl$ was used for reducing ${IO_3}^-$ to $I_2$.The optimum acidity of the dissolved solution and the added of $NH_2OH{\cdot}HCl$ were 2.5 M and more than $1.5{\times}10^{-3}mole$, respectively. The recovery of iodide by ion chromatography was $82.8{\pm}4.1%$ and the total yield was corrected by gamma spectrometery using $^{131}I$ as a tracer.