• 대한방사선방어학회:학술대회논문집
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• 2010.04a
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• pp.198-199
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• 2010

• 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.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.48-55
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• 1989

In this study, simple functional forms which could predict decay heat are referred to and modified in order to analyse more easily long-term behavior of decay heat generated from domestic PWR and CANDU spent fuel. To reduce the difference between the predicted data by functional forms and ORIGEN 2 results and to predict the decay heat under the important parameter(s), sensitivity analysis is performed. By introducing the identified hey parameter, turnup, into the functional forms, the decay heat of spent fuels within a limited rangs of cooling time(3~500 years) becomes predictable for various turnup rates. The predicted decay heat of spent fuels with representative turnup rates such as 33, 37 and 40 GWD/MTU by the functional forms is in so good agreement with ORIGEN 2 results within $\pm$10% difference over the cooling time from 1 to 10$^{5}$ years that the functional forms presented here may be used for engineering purposes such as the thermal design and assessment of the facilities associated with spent fuel management.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.268-273
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• 1997

사용후핵연료는 장기간 강한 방사선과 붕괴열이 방출된다. 따라서 사용후핵연료를 안전하게 운반하기 위하여 수송용기는 방사선차폐의 건전성, 격납경계의 유지 및 내부 붕괴열의 적절한 제거 등의 설계기준을 만족하도록 설계되어야 한다. 본 연구에서는 28개의 PWR 사용후핵연료집합체를 운반할 수 있는 KSC-28 수송용기의 적절한 열전달 특성을 갖는 copper 냉각핀 및 aluminum 전열판을 설정하였다. 또한, 정상수송조건 및 화재사고조건에 대한 열전달해석을 수행하여 수송용기의 열적 건전성을 평가하였고 여기에서 얻어진 온도를 열하중으로 고려하여 열응력해석을 수행함으로써 수송용기의 온도변화에 따른 구조적 건전성을 평가하였다.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.8-19
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• 1993

The natural convection model of the consolidated system has been developed to make sure the removal of decay heat generated in the spent fuel for the loss of forced cooling accident. The numerical technique employed was based on the ADI scheme. The calculation of heat generation rate in the spent fuel was peformed by the ANS-79 decay heat model, and the nonuniform surface heat flux is assumed with a chopped sine curve for the conservative decay heat generation input. The sensitivity study was performed to examine the possibility of the pool bulk boiling by varying the various parameters, i.e. inter-fuel spacing ratio, heat generation power, and radius of the fuel rod. The application results of this model show that the natural circulation flow through compacted spent fuel bundles enables the pool temperature to control in a safe and effective manner, after the required cooling time. The corresponding acceptance criteria of the cooling time for rearranging the spent fuel rods were also found.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.201-206
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• 1996

건식가공(Dry Process)이 사용전,후 DUPIC 핵연료의 붕괴열(Decay Heat), Hazard Index, 조사선량률(Dose Rate) 등에 미치는 영향을 계산하고, 그 원인을 분석하였다. DUPIC 사용방안으로 표준 연소도(35,000 MWD/MTU)의 경우와 장주기 연소도(50,000 MWD/MTU)의 경우를 고려하여 계산하였으며, DUPIC핵연료는 20년 냉각후 가공하는 것을 기준으로 하였다. 또한 DUPIC핵연료 장전시 고려할 수 있도록 사용전 DUPIC 핵연료에 대한 계산을 핵연료 집합체(Bundle) 단위로 하였다. 조사선량과 붕괴열은 건식가공에 상당히 민감한 반응을 보였고 이는 주로 Cs의 제기에 의한 것이다.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.19-31
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• 1994

The objective of this study is to conduct a thermal-hydraulic analysis on the spent fuel pool and to evaluate a parametric effect for the thermal-hydraulic analysis of spent fuel pool. The selected parameters are the Reynolds Number and the gap flow through the oater gap between fuel cell and fuel bundle. The simplified flow network for a path of fuel cells is used to analyze the natural circulation phenomenon. In the flow network analysis, the pressure drop for each assembly from the entrance of the fuel rack to the exit of the fuel assembly is balanced by the driving head due to the density difference between the pool fluid and the average fluid in each spent fuel assembly. The governing equations ore developed using this relation. But, since the parameters(flow rate, pressure loss coefficient, decay heat, density)are coupled each other, iteration method is used to obtain the solution. For the analysis of the YGN 3&4 spent fuel rack, 12 channels are considered and the inputs such as decay heat and pressure loss coefficient are determined conservatively. The results show the thermal-hydraulic characteristics(void fraction, density, boiling height)of the YGN 3&4 spent fuel rack. There occurs small amount of boiling in the cells. Fuel cladding temperature is lower than 343.3$^{\circ}C$. The evaluation of parametric effect indicates that flow resistances by geometric effect are very sensitive to Reynolds number in the transition region and the gap flow is negligible because of the larger flow resistance in the gap flow path than in the fuel bundle.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.161-162
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• 2015

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.219-220
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• 2016

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.73-78
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• 2009

The decay heat of Cs and Sr contained in a LiCl waste salt, generated from an electrolytic reduction process in pyroprocessing of spent nuclear fuel, has been calculated. The calculation has been carried out under some assumptions that most of the LiCl waste is purified and recycled to main process, and the residual is fabricated to make a waste form. As a result, the decay heat from daughter nuclides such as Ba and Y seems to be maximum 4.6 times higher than that from their parent nuclides such as Cs and Sr. The thermal release from Cs and Sr in the LiCl waste is the maximum around the first one month, so an cooling system operation for some time at the beginning would be suggested to control a rapid increase in the temperature of the LiCl waste salt.