• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.217-224
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• 2008

In deep geological disposal system, the integrity of a disposal canister having spent fuels is very important factor to assure the safety of the repository system. This disposal canister is one element of the engineered barriers to isolate and to delay the radioactivity release from human beings and the environment for a long time so that the toxicity does not affect the environment. The main requirement in designing the deep geological disposal system is to keep the buffer temperature below 100$^{\circ}C$ by the decay heat from the spent fuels in the canister in order to maintain the integrity of the buffer material. Also, the disposal canister can endure the hydraulic pressure in the depth of 500 m and the swelling pressure of the bentonite as a buffer. In this study, new concept of the disposal canister for the CANDU spent fuels which were considered to be disposed without any treatment was developed and the thermal stability and the structural integrity of the canister were analysed. The result of the thermal analysis showed that the temperature of the buffer was 88.9$^{\circ}C$ when 37 years have passed after emplacement of the canister and the spacings of the disposal tunnel and the deposition holes were 40 m and 3 m, respectively. In the case of structural analysis, the result showed that the safety factors of the normal and the extreme environment were 2.9 and 1.33, respectively. So, these results reveal that the canister meets the thermal and the structural requirements in the deep geological disposal system.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.446-454
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• 2021

Critical heat flux (CHF) has traditionally been evaluated using look-up tables or empirical correlations for nuclear power plants. However, under complex moving condition, it is necessary to reconsider the CHF characteristics since the conventional CHF prediction methods would no longer be applicable. In this paper, the additional forces caused by motions have been added to the annular film dryout (AFD) mechanistic model to investigate the effect of moving condition on CHF. Moreover, a theoretical model of the natural circulation loop with additional forces is established to reflect the natural circulation characteristics of the loop system. By coupling the system loop with the AFD mechanistic model, a CHF prediction program called NACOM for natural circulation loop under moving condition is developed. The effects of three operating conditions, namely stationary, inclination and rolling, on the CHF of the loop are then analyzed. It can be clearly seen that the moving condition has an adverse effect on the CHF in the natural circulation system. For the calculation parameters in this paper, the CHF can be reduced by 25% compared with the static value, which indicates that it is important to consider the effects of moving condition to retain adequate safety margin in subsequent thermal-hydraulic designs.

• Journal of the Korean Society of Marine Environment & Safety
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• v.20 no.3
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• pp.304-311
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• 2014

In this study, to apply hydrogen energy to ship engine and to generate effective hydrogen production, we investigated the effects of high temperature $H_2SO_4$ feed rate and cooling water rate to pump parts with fixed frequency needed to reciprocate motion and a simulation was conducted at each condition. In the fixed frequency and cooling water inlet flow rate of 0.5 Hz and 3.9 kg/s, we changed the high temperature $H_2SO_4$ flow rate to 47.46 kg/s (it is 105 % of 45.2 kg/s), 49.72 kg/s (110 %), and 51.98 kg/s (115 %). Also, at 0.5 Hz and 45.2 kg/s of frequency and high temperature $H_2SO_4$ flow, the thermal hydraulic analysis was performed at the condition of 95 % (3.705 kg/s), 90 % (3.51 kg/s), and 85 % (3.315 kg/s). In overall simulation cases, the physical properties of materials are more influential to the temperature increase in the pump part rather than the changes on the feed rate of high temperature $H_2SO_4$ and cooling water. A continuous operation of pump was also capable even if the excess feed of high temperature $H_2SO_4$ of about 15 % or the less feed of cooling water of about 15 % were performed, respectively. When the increasing feed of high temperature $H_2SO_4$ of up to 5 %, 10 %, and 15 % were compared with base flow (45.2 kg/s), the deviation of time period rose to a certain temperature and ranged from 0 to 4.5 s in the same position (same material). In case of cooling water, the deviation of time period rose to a certain temperature and ranged from 0 to 5.9 s according to the decreasing feed changes of cooling water at 5 %, 10 %, and 15 % compared to a base flow (3.9 kg/s). Finally, the additional researches related to the two different materials (Teflon and STS for Pitch and End-plate), which are concerned about the effects of temperature changes to the parts contacting different materials, are needed, and we have a plan to conduct a follow-up study.

• Journal of Energy Engineering
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• v.25 no.2
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• pp.1-20
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• 2016

A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.237-245
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• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.