• Nuclear Engineering and Technology
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• 제32권4호
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• pp.309-315
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• 2000

The changes of Np oxidation state in nitric acid and the effect of nitrous acid on the oxidation state were analyzed by spectrophotometry, solvent extraction, and electrochemical methods. An enhancement of Np extraction to 30 vol.% TBP was carried out through adjustment of Np oxidation state by using a glassy carbon fiber column electrode system. The information of electrolytic behavior of nitric acid was important because the nitrous acid affecting the Np redox reaction was generated during the electrolytic adjustment of the Np oxidation state. The Np solution used in this work consisted of Np（V) and Np（Ⅵ）without （IV）. The composition of Np（V) in the range of 0.5M -5.5 M nitric acid was 32% ~ 19%. The electrolytic oxidation of Np（V) to Np（Ⅵ）in the solution enhanced Np extraction efficiency about five times higher than the case without the electrolytic oxidation. It was confirmed that the nitrous acid of less than about 10-5 M acted as a catalyst to accelerate the chemical oxidation reaction of Np（V) to Np（Ⅵ）.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.415-420
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• 2013

A determination method for $^{237}Np$ in spent nuclear fuel samples was developed using an isotope dilution method with $^{239}Np$ as a spike. In this method, inductively coupled plasma mass spectrometry (ICP-MS) was taken for the $^{237}Np$ instead of the previously used alpha spectrometry. $^{237}Np$ and $^{239}Np$ were measured by ICP-MS and gamma spectrometry, respectively. The recovery yield of $^{237}Np$ in synthetic samples was $95.9{\pm}9.7$% (1S, n=4). The $^{237}Np$ contents in the spent fuel samples were 0.15, 0.25, and $1.06{\mu}g/mgU$ and these values were compared with those from ORIGEN-2 code. A fairly good agreement between the measurements (m) and calculations (c) was obtained, giving ratios (m/c) of 0.93, 1.12 and 1.25 for the three PWR spent fuel samples with burnups of 16.7, 19.0, and 55.9 GWd/MtU, respectively.

• 방사성폐기물학회지
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• 제6권3호
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• pp.163-170
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• 2008

30 % TBP/NDD-2 M $HNO_3$ (0.005 M $K_2Cr_2O_7$ 함유) 계에 의해 산화 추출된 Np의 유기상을 대상으로 n-butyraldehyde (NBA)에 의한 Np의 환원 역추출을 고찰하였다. Np의 역추출은 NBA의 농도 증가에 따라, 역추출 수용상 내 질산농도 감소에 따라, 그리고 반응 온도 감소에 따라 증가하였으며, 이때 겉보기 환원 역추출 속도식은 $-d[Np]_{Org.}/dt$ =1,524 exp(-2,906/T) $[NBA]^{0.91}\;[H^+]^{-0.92}[Np]_{Org.}$. 이었다. 1.04M NBA 및 2M $HNO_3$에서 Np 및 U의 역추출률은 각각 70.1 % 및 7.1 % 이며 이때 분리계수(($=D_U/D_{Np}$)는 30.4 정도로, TBP-$HNO_3$ 계에 의해 공추출된 Np과 U은 NBA에 의해 효과적으로 상호 분리할 수 있음을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.