• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3723-3740
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• 2021

The harsh conditions in the reactor affect the thermal and mechanical performance of the fuel plate heavily. Some in-pile behaviors, like fission-induced swelling, can cause a large deformation of fuel plate at very high burnup, which may even disturb the flow of coolant. In this research, the emphasis is put on the thermal expansion, fission-induced swelling, interaction layer (IL) growth, creep of the fuel meat, and plasticity of the cladding for the U₃Si₂/Al dispersion fuel plate. A detailed model of the fuel meat swelling is developed. Taking these in-pile behaviors into consideration, the three-dimensional large deformation incremental constitutive relations and stress update algorithms have been developed to study its thermal-mechanical performance under normal conditions using Abaqus. Results have shown that IL can effectively decrease the thermal conductivity of fuel meat. The high Mises stress region mainly locates at the interface between fuel meat and cladding, especially around the side edge of the interface. With irradiation time increasing, the stress in the fuel plate gets larger resulting from the growth of fuel meat swelling but then decreases under the effect of creep deformation. For the cladding, plasticity deformation does not occur within the irradiation time.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.129-130
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• 2018

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.58-59
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• 2018

• 방사성폐기물학회지
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• 제20권2호
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• pp.255-258
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• 2022

Three government bodies, that is, the Ministry of Science and ICT (MSIT), Ministry of Trade, Industry, and Energy (MOTIE), and Nuclear Safety and Security (NSSC), jointly established the Institute for Korea Spent Nuclear Fuel (iKSNF) in December 2020 to secure the management technologies for spent nuclear fuel (SNF). The objective of iKSNF is to successfully conduct the long-term research and development program of the 「Development of Core Technologies to Ensure Safety of Spent Nuclear Fuel Storage and Disposal System」. Our program, known as the first multi-ministry program in the nuclear field of Korea, mainly focuses on developing core technologies required for the long-term management of SNF, including those for safe storage and deep geological disposal of SNF. The program comprises three subprograms and seven key projects covering the storage, disposal, and regulatory sectors of SNF management. Our program will last from 2021 through 2029, with a budget of approximately four billion USD sponsored by MSIT, MOTIE, and NSSC.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3071-3079
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• 2023

Helium production in the nuclear fuel matrix during irradiation plays a critical role in the design and performance of Gen-IV reactor fuel, as it represents a life-limiting factor for the operation of fuel pins. In this work, a surrogate model for the helium production rate in fast reactor MOX fuels is developed, targeting its inclusion in engineering tools such as fuel performance codes. This surrogate model is based on synthetic datasets obtained via the SCIANTIX burnup module. Such datasets are generated using Latin hypercube sampling to cover the range of input parameters (e.g., fuel initial composition, fission rate density, and irradiation time) and exploiting the low computation requirement of the burnup module itself. The surrogate model is verified against the SCIANTIX burnup module results for helium production with satisfactory performance.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.155-156
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• 2017

In this study, the structural integrity of the spent nuclear fuel assemblies was evaluated by carrying out a 0.3 m drop impact analysis, one of the normal transportation conditions of the nuclear fuel assemblies. For this purpose, the spent nuclear fuel assembly was modeled in detail as beam elements, and a coupled model for impact analysis was developed by inserting the modeled nuclear fuel assemblies into a cask.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.178-187
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• 2021

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.

• Nuclear Engineering and Technology
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• 제50권2호
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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.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1063-1070
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• 2017

This paper presents a method for forecasting future uranium prices that is used as input data to calculate the uranium cost, which is a rational key cost driver of the nuclear fuel cycle cost. In other words, the statistical autoregressive integrated moving average (ARIMA) model and existing engineering cost estimation method, the so-called escalation rate model, were subjected to a comparative analysis. When the uranium price was forecasted in 2015, the margin of error of the ARIMA model forecasting was calculated and found to be 5.4%, whereas the escalation rate model was found to have a margin of error of 7.32%. Thus, it was verified that the ARIMA model is more suitable than the escalation rate model at decreasing uncertainty in nuclear fuel cycle cost calculation.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2604-2612
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• 2023

Spent nuclear fuel and long-lived radioactive waste must be carefully handled before disposing them off to a geological repository. After the pre-storage period in water pools, spent nuclear fuel is stored in casks, which are widely used for interim storage. Interim storage in casks is very important part in the whole cycle of nuclear energy generation. This paper presents the results of the numerical study that was performed to evaluate the thermal behavior of a metal-concrete CONSTOR RBMK-1500 cask loaded with spent nuclear fuel and placed in an open type interim storage facility which is under fire conditions (steady-state, fire, post-fire). The modelling was performed using the ANSYS Fluent code. Also, a local sensitivity analysis of thermal parameters on temperature variation was performed. The analysis demonstrated that the maximum increase in the fuel load temperatures is about 10 ℃ and 8 ℃ for 30 min 800 ℃ and 60 min 600 ℃ fires respectively. Therefore, during the fire and the post-fire periods, the fuel load temperatures did not exceed the 300 ℃ limiting temperature set for an RBMK SNF cladding for long-term storage. This ensures that fire accident does not cause overheating of fuel rods in a cask.