• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.98-104
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• 1996

핵분열 생성물 방출량을 계산하는 모델들에 대한 비교 분석을 위해 GAPCON-THERMAL-2 Revision 2 (GT2R2) 코드를 이용하여 Beyer-Hann , Beyer-Hann with NRC High Burnup Correction, ANS5.4와 Modified ANS5.4 핵분열 생성물 방출 모델들을, RISO-M2-2C 핵연료봉의 실험결과와 비교하였다. Beyer-Hann 모델은 실험결과보다 낮게 예측한반면 ANS5.4 모델은 실험결과 보다 높게 예측하였다. 한편 NRC High Burnup Correction을 한 Beyer-Hann 모텔과Modified ANS5.4 모델은 실험 결과와 비슷한 방출비를 예측하였다. 이러한 결과를 확인하기 위해 국부적인 핵연료 온도와 연소도를 검토한 결과 ANS5.4 모델이 .Modified ANS5.4 모델보다 온도와 연소도에 따라 더 민감한 반응을 보이고 있으며, Beyer-Hann 모텔은 연소도 영향이 없이 각 온도 영역에서 일정하였고, Beyer-Hann with NRC High Burnup Correction 모델은 20,000MWd/MTU 연소도 이상영역에서 연소도 영향을 보이고 있다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.139-144
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• 1996

핵분열 생성물 방출량을 계산하는 모델들에 대한 비교 분석을 위해 GAPCON-THERMAL-2 Revision 2 (GT2R2) 코드를 이용하여 Beyer-Hann , Beyer-Hann with NRC High Burnup Correction, ANS5.4와 Modified ANS5.4 핵분열 생성물 방출 모델들을, RISO-M2-2C 핵연료봉의 실험결과와 비교하였다. Beyer-Hann 모델은 실험결과보다 낮게 예측한반면 ANS5.4 모델은 실험결과 보다 높게 예측하였다. 한편 NRC High Burnup Correction을 한 Beyer-Hann 모델과 Modified ANS5.4 모델은 실험 결과와 비슷한 방출비를 예측하였다. 이러한 결과를 확인하기 위해 국부적인 핵연료 온도와 연소도를 검토한 결과 ANS5.4 모델이 Modified ANS5.4 모델보다 온도와 연소도에 따라 더 민감한 반응을 보이고 있으며, Beyer-Hann 모델은 연소도 영향이 없이 각 온도 영역에서 일정하였고, Beyer-Hann with NRC High Burnup Correction 모델은 20,000MWd/MTU 연소도 이상영역에서 연소도 영향을 보이고 있다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.125-133
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• 2004

Fission products of DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel, irradiated in HANARO research reactor with 61 ㎾/m of maximum linear power and 1,770 ㎿d/tU of average burn-up, was characterized by EPMA(Electron Probe Micro Analyzer). In order to find accurate characterization, the analysis results by EPMA of fresh simulated DUPIC fuel containing fission products as chemicals were compared with that of wet chemical analysis. The metallic precipitates observed at the center of the fresh simulated DUPIC fuel were about 1 $\mu\textrm{m}$ in size and their major components by EPMA were Mo-53.89 at.%, Ru-37.40 at.%, and Pd+Rh-8.71 at.%. Established procedure through the fresh simulated DUPIC fuel was applied to the irradiated DUPIC fuel. Observed size of metallic precipitates were 2∼2.5 $\mu\textrm{m}$ and their compositions were Mo-47.34 at.%, Ru-46 at.%, and Pd+Rh-6.65 at.%. What are uncommon things for this experiment, special treatment for improving the conductivity was attempted to the specimen and the conditions of exact irradiation of electron beam to small metallic precipitate were suggested.

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.40-51
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• 1992

One of the important irradiation performance characteristics of the silicide dispersion fuel element in research reactors is the diameteral increase resulting from fuel swelling. This paper, will attempt to develop a physical model for the fuel swelling, DFSWELL, by analyzing the basic irradiation behaviours and some experimental evidences. From the experimental evidences, it was shown that the volume changes in irradiated U$_3$Si-Al were strongly dependent on temperature and fission rate. The quantitative-amount of swelling for silicide fuel is estimated by considering temperature, fission rate, solid fission product build-up and gas bubble behavior. The swelling for the silicide fuel is comprised of three major components : i ) a volume change due to the formation of an interfacial layer between the fuel particle and matrix. ii ) a volume change due to the accumulation of gas bubble nucleation iii ) a volume change due to the accumulation of solid fission products The DFSWELL model which takes into account the above three major physical components predicts well the absolute magnitude of silicide fuel swelling in accordance with the power histories in comparison with the experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.362.1-362
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• 2001

• Journal of Energy Engineering
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• v.23 no.3
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• pp.186-190
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• 2014

Nuclear Power Plants require the control ability of chemical condition (pH) because pH control during transient accident such as LOCA makes an able the fission product removal capability to be maintained, stress corrosion cracking of stainless steel equipment to be prevented and the production of hydrogen by aluminum and zinc to be minimized. An NPP is designed to control the pH of containment spray and sump coolant using the spray additives 30% NaOH in the event of loss of coolant accident. In this paper, the pH of sump coolant of an NPP during LOCA was analyzed and the fission products removal constant and decontamination factor were calculated according to Standard Review Plan 6.5.2 related to spray chemical conditions of pH. The calculated pH value of recirculation mode using the computer code corresponds to 8.09~9.67, which meets the chemical environment regulation requirements. The fission product removal capability caused by containment spray system is performed to provide input to radiation analysis.

• Nuclear Engineering and Technology
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• v.16 no.3
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• pp.169-179
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• 1984

An efficient procedure of evaluating the fuel cladding failures occurring in the normal operations of typical PWR's has been investigated through the analysis of fission product(FP) activities in the reactor coolant using an analytical model, FIPREL code. Performed by this code is an extensive study on the sensivities of FP activities to such physical parameters as enrichment, turnup, and operation temperature of failed fuel rod as well as the effective failure size quantified in terms of the magnitude of gap release coefficient. The results of study are generally in agreement with those by PROFIP method. In the presence of tramp uranium the portion of activities released from failed rod is separated by an iterative calculation based on the activity ratios of fission nuclides chemically more stable than iodines. Obtained are the linear power density and the number of failed rods, the effective failure size, and the mass of tramp uranium. The operation experiences of 4 cycles of Kori Unit 1 are analyzed and the results show that the model is highly reliable for the survey and evaluation of fuel rod conditions during reactor operations.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.213-213
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• 2002

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.05a
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• pp.173-174
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• 2011

• Journal of the Korean Chemical Society
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• v.45 no.4
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• pp.304-311
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• 2001

A study has been carried out on the extraction chromatographic separation of fission products from spent pressurized water reactor (PWR) fuels for inductively coupled plasma atomic emission spectrometric analysis. Impregnation capacity of tri-n-butyl phosphate (TBP), which is well known as an extractant in the field of uranium separation from various nuclear grade materials, on Amberlite XAD polymeric macroporous support materials was measured. Amberlite XAD-16 of which the surface area is the highest was selected as a support material because its TBP impregnation capacity was the largest in Amberlite XADs. Sorption behaviour of this TBP impregnated resin was investigated for the fission product elements using acidic solutions simulated for dissolver solutions of spent PWR fuels. The parameters affecting the performance of the separation system were optimized. The fission product elements studied excluding Pd and Ru were quantitatively recovered with the precision of less than 3.1%.