• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.95-102
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• 2006

Alpha spectrometry was studied for the determination of $^{237}Np$ in spent nuclear fuel samples. The optimum condition for the electrodeposition of $^{237}Np$ was obtained as follows : for $1{\sim}1.5$ hour of deposition time, at the current intensity of $1.2{\sim}1.5$ A and at sodium sulfate electrolyte without organic additive. The deposition yield and its reproducibility on $^{237}Np$ was decreased as the amount of $^{237}Np$ decreased from 4.16 Bq down to 0.0264 Bq(1ng). The recovery yield of $^{237}Np$ determined by alpha spectrometry after separation in synthetic solution was $98.8{\pm}5.1%$(n=4). The contents of $^{237}Np$ in spent nuclear fuel samples were determined and the result showed an agreement within 10% of a difference between the measurement and the calculation.

• Nuclear Engineering and Technology
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• v.45 no.3
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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.

• Applied Chemistry for Engineering
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• v.7 no.1
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• pp.153-161
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• 1996

Several main nuclides($^{241}Am$, $^{152}Eu$ and $^{237}Np$) in radioactive waste solution were selected and examined to mutual separation with di-(2-ethylhexyl) phosphoric acid by solvent extraction technique. $^{237}Np$ was extracted more than 99.9% adding the $H_2O_2$ that was a good reductant for the oxidation state control of $^{237}Np$. $^{241}Am$, $^{152}Eu$ and $^{237}Np$ could be fairly well separated one another during the different sequence stripping stages, but about 7~9.6% of the other nuclides were still remained for the $^{241}Am$ stripping solution. This result shows that the product of $^{152}Eu$ and $^{237}Np$ was good, but $^{241}Am$ may be needed to further purification process. It was also discussed on the cause of the third phase formation phenomenon that was found in the solvent regeneration.

• Journal of the Korean Society of Radiology
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• v.5 no.4
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• pp.217-221
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• 2011

$^{237}Np$ is very important material in the fission products of nuclear reactors. Resonance integral(RI) tests of this material is necessary to check between the experiments and the evaluated data. Such feedback to the evaluated data is very important to correct data and improve of codes. The RI for the $^{237}Np(n,{\gamma})^{238}Np$ reaction were measured by using the 46-MeV electron linear accelerator (linac) at the Research Reactor Institute, Kyoto University (KURRI). The measurement was performed in the energy region from 0.005 eV and 10 keV. RI obtained as 804.7 barns, compared with those of the evaluated data in JENDL-4.0 and Mughabghab.

• Nuclear Engineering and Technology
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• v.13 no.4
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• pp.207-220
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• 1981

The Np-237 fission cross-sections up to 20 MeV incident neutron energy are calculated by means of the computer code STAPRE- a statistical model code with consideration of preequilibrium decay. The higher chance fissions up to third compound nucleus are taken into account, and the main input parameters in the treatment of fission under consideration of a double-humped fission barrier are carefully adjusted, so that the current trend of experimental data can be fitted within an apparent deviation of about 10% throughout the entire energy range. Results are presented in the form of point-wise cross-section values, and also in the form of graph to demonstrate the shape agreement.

• Nuclear Engineering and Technology
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• v.3 no.2
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• pp.77-97
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• 1971

The experimental fission cross-section data of U-238, Np-237 and Th-232, published up to the end of 1970, are reviewed and analyzed between their respective thresholds and 20.0 MeV. The results of a statistical analysis of the available data, performed with a weighted Least-squares Orthogonal Polynomial Pitting computer programme are presented in the form of point-wise cross-section values together with their uncertainties, and in the form of graphs of the fitted curves with an indication of a region of 95% statistical confidence level. An estimate of the fission spectrum weighted average cross-sections and their respective uncertainties is also given.

• 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.33 no.2
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• pp.231-240
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• 2001

When it is assumed that PWR, CANDU and DUPIC spent fuels are disposed of in deep geological repository, consequent annual individual doses are calculated, and it is shown that doses meet the regulatory limit. From these results, the hazardous radionuclides applicable to partitioning and transmutation are selected. These selected radionuclides such as Tc-99, Ⅰ-129, Cs-135 and Np-237 are then reviewed in terms of partitioning and transmutation. Separation of I-129, Np-237 and Tc-99 from spent fuels is considered desirable, and transmutation of these radionuclides results in remarkable hazard reduction. However, it is concluded that separation and transmutation of Cs-135 may be ineffective although it is classified into a hazardous radionuclide.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.182-190
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• 2004

An unique extraction chromatographic column (TOA: Tri-n-octylamine on Teflon powder) with a two-stage sample loading was prepared to separate Np and Pu from the environmental matrix. Np and Pu were efficiently retained in 4 M $HNO_3$ medium on the column and easily eluted with 0.02 M oxalic acid in 0.16 M HNO$_3$at $95^{\circ}C$. The separated solutions were free from most of the matrix elements and were aspirated into the ICP-MS directly. The decontamination factor for 238U is more than 104. The instrumental detection limit for 237Np was 0.46 pg mL-l ($1.2{\times}10_{-5}$Bq mL-l), and for 239Pu was 0.48 pg mL-l ($1.1{\times}10_{-3}$ Bq mL-l). The feasibility for the determination of both elements was proved by analysing IAEA-135 reference samples, the measured values agreed with the recommended reference value.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.252-253
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• 2005