• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2723-2733
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• 2023

Transuranic nuclides (such as ²³⁸Pu, ²⁵²Cf, ²⁴⁹Bk, etc.) have a wide range of application in industry, medicine, agriculture, and other fields. However, due to the complex conversion chain and remarkable fission losses in the process of transuranic nuclides production, the generation amounts are extremely low. High flux reactor with high neutron flux and flexible irradiation channels, is regarded as the promising candidate for producing transuranic nuclides. It is of great significance to increase the conversion ratio of transuranic nuclides, resulting in higher efficiency and better economy. In this paper, we perform an optimization design evaluation of producing transuranic nuclides in high flux reactor, which includes optimization design of irradiation target and influence study of reactor core loading. It is demonstrated that the production rate increases with appropriately determined target material and target structure. The target loading scheme in the irradiation channel also has a significant influence on the production of transuranic nuclides.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.99-106
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• 2009

The contents of transuranic elements ($^{237}Np$, $^{238}Pu$, $^{239}Pu$, $^{240}Pu$, $^{241}Am$, $^{244}Cm$, and $^{242}Cm$) in high-burnup spent fuel samples ($35.6{\sim}53.9\;GWd/MtU$) were determined by alpha spectrometry. Anion exchange chromatography and diethylhexyl phosphoric acid extraction chromatography were applied for the separation of these elements from the uranium matrix. The measured values of the nuclides were compared with ORIGEN-2 calculations. For plutonium, the measurements were higher than the calculations by about $2.6{\sim}32.7%$ on average according to each isotope, and those for americium and curium were also higher by about $35.9{\sim}63.1%$. However, for $^{237}Np$, the measurements were lower by about 52% on average for the samples.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1060-1067
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• 2018

A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial $k_{eff}$ level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially $^{241}Am$ is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.351-357
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• 2003

Transuranic elements such as Pu, Am and Cm in synthetic solution of spent nuclear fuel samples were determined by electrodeposition followed by alpha-spectrometry after separation using anion exchange and extraction chromatography in order to determine the transuranic elements in radwaste samples from nuclear power plants. Plutonium was separated by 12M HC1-0.1M HI as an eluent on anion exchange column. As a second step Am and Cm were separated in a group by DTPA-Lactic acid as the eluent on HDEHP coated column. The nuclides of $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ separated were determined by alpha-spectrometry after electrodeposition in 0.1M $NaHSo_4$-0.53M $Na_2SO_4$buffer solution as an electrolyte. The recovery yields of $^{239}Pu$, $^{241}Am$및 $^{244}Cm$ were 83.8%, 85.2% and 86.3%, respectively, from the synthetic solution containing uranium and non-radioactive metal elements.

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.459-482
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• 2006

A geochemical approach to the long-term safety of waste disposal is discussed in connection with the significance of actinides, which shall deliver the major radioactivity inventory subsequent to the relatively short-term decay of fission products. Every power reactor generates transuranic (TRU) elements: plutonium and minor actinides (Np, Am, Cm), which consist chiefly of long-lived nuclides emitting alpha radiation. The amount of TRU actinides generated in a fuel life period is found to be relatively small (about 1 wt% or less in spent fuel) but their radioactivity persists many hundred thousands years. Geological confinement of waste containing TRU actinides demands, as a result, fundamental knowledge on the geochemical behavior of actinides in the repository environment for a long period of time. Appraisal of the scientific progress in this subject area is the main objective of the present paper. Following the introductory discussion on natural radioactivities, the nuclear fuel cycle is briefly brought up with reference to actinide generation and waste disposal. As the long-term disposal safety concerns inevitably with actinides, the significance of the aquatic actinide chemistry is summarized in two parts: the fundamental properties relevant to their aquatic behavior and the geochemical reactions in nanoscopic scale. The constrained space of writing allows discussion on some examples only, for which topics of the primary concern are selected, e.g. apparent solubility and colloid generation, colloid-facilitated migration, notable speciation of such processes, etc. Discussion is summed up to end with how to make a geochemical approach available for the long-term disposal safety of nuclear waste or for the performance assessment (PA) as known generally.