• Nuclear Engineering and Technology
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• v.39 no.5
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• pp.673-682
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• 2007

The contents of transuranic elements in high-burnup spent fuel samples were determined. The activity amounts of $^{238}Pu,\;^{239}Pu,\;^{240}Pu,\;^{241}Am,\;^{244}Cm\;and\;^{242}Cm$ were measured by alpha spectrometry using $^{242}Pu\;and\;^{243}Am$ as tracers, respectively. A spike addition method for $^{237}Np$ was established by an alpha and gamma spectrometry using $^{239}Np$ as a spike after the optimum conditions for the measurements of $^{237}Np\;and\;^{239}Np$, respectively, were obtained. A separation system using anion exchange chromatography and diethylhexylphosphoric acid extraction chromatography was applied for the separation of these elements. This method was applied to high-burnup spent nuclear fuel samples $(40{\sim}60GWD/MTU)$. The contents of the transuranic elements were compared with those by ORIGEN-2 code. Measurements and the calculations of the contents of the plutonium isotopes $^{238}Pu,\;^{239}Pu\;and\;^{240}Pu$ agreed to within 10% on average. The contents of $^{237}Np$ agreed to within approximately 5% except for one instance of a calculation, while those of $^{241}Am,\;^{244}Cm\;and\;^{242}Cm$ showed higher values by approximately 19%, 35% and 14% on average, respectively, compared to the calculations according to the burnup.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.79-88
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• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.179-186
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• 2009

This paper presents a thermal creep test under internal pressure and post-test characterization performed on high burnup (68 MWd/kgU) ZIRLO. This research has been done by the CSN, ENRESA, and ENUSA in order to investigate the behavior of advanced cladding materials in contemporary PWRs at higher burnup under dry cask storage conditions. Also, to investigate the hydride reorientation, the cool-down of the samples after the test has been done in a coordinated manner with the internal pressure. The creep results obtained are consistent with the expected behavior from reference CWSR material, Zr-4. During the test, the material retained significant ductility: one specimen leaked during the test at an engineering strain of the tube section of 17%; remarkably, the crack closed due to de-pressurization. Although significant hydride reorientation occurred during the cool-down under pressure, no specimen failed during the cool-down.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.294-302
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• 1999

The influence of source term release parameters on offsite health effects was examined for the YGN 3&4 nuclear power plants. The release parameters considered in this study are release height, heat content, and release time. The effects of core inventory change as a function of fuel burnup was also examined. The health effects by the change of release parameters are early fatalities, cancer fatalities, and early fatality distance. The results showed that early fatalities and early fatality distance decrease as release height increases, although it does not have significant influence on cancer fatalities. The values of both early and late health effects decrease as heat content increases. As release time increases, health consequence shows maximum value in 2 hours of release time and then decreases rapidly. As fuel burnup increases, early fatalities decrease rapidly, while cancer fatalities increase rapidly. Both cases show little variation afterward. Early fatality distance is almost same in all fuel turnup history. The information obtained through this research is very useful in developing strategies for reducing offsite consequences when combined with the influence of weather conditions on offsite risks.

• 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.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.407-412
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• 2014

In this research, we investigated the burnup characteristics and the conversion of fertile $^{232}Th$ into fissile $^{233}U$ in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning $^{232}Th$ fuel (fuel pin 1) and $^{233}U$ fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.674-680
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• 2020

We used MCNP6 computer code to model HTR-10 core reactor. We used two types of fuel; UO₂ and (Th+Pu)O₂ mixture. We determined the critical height at which the reactor approached criticality in both two cases. The neutronic and burnup parameters were investigated. The results indicated that the core fueled with mixed (Th+Pu)O₂, achieved about 24% higher fuel cycle length than the UO₂ case. It also enhanced safeguard security by burning Pu isotopes. The results were compared with previously published papers and good agreements were found.

• Nuclear Engineering and Technology
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• v.35 no.4
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• pp.336-346
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• 2003

This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction.

• Nuclear Engineering and Technology
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• v.55 no.8
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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.