• Title/Summary/Keyword: uranium conversion plant

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A Study on the Natural Uranium Contamination Measuring Technology (천연우라늄 오염에 관한 방사선/능 측정기술 연구)

  • 정운수;홍상범;서범경;박진호;조용우;조성원;이정민
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2004.06a
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    • pp.407-417
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    • 2004

  • This study is to verify radiation detection method by using $\alpha$-spectroscopy and ${\gamma}$-spectroscopy for concretes and components which will be generated during the decommissioning of the uranium conversion plant. Components and inside walls of the building were contaminated with natural uranium materials. Some parts of the stainless steel pipes and concretes of the walls were sampled and analyzed their alpha and gamma activities respectively. Alpha and gamma activities are well matched each other in the range of high activity region to 0.01 Bq/g and gamma activities are over estimated comparing alpha activities corresponded in below 0.005 Bq/g region for the natural uranium of AUC sample. The $^{238}U$ originated from natural products of conversion process could be distinguished by measuring $^{214}Pb$ or $^{214}Bi$ and $^{234}Th$ or $^{234m}Pa$. Uranium contaminations mainly are in the wall surface of the plant. Decontamination process of generating wastes which can be reached tp background level gamma activities measured by gamma spectroscopy can also be used to conservative assessment data.

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Electrochemical Decontamination of Metallic Wastes Contaminated with Uranium Compounds in a Neutral Salt Electrolyte

  • Park, W. K.;Y. M. Yang;C. H. Jung;H. J. Won;W. Z. Oh;Park, J. H.
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2003.11a
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    • pp.689-695
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    • 2003

  • Electrochemical decontamination process has been applied for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds such as $UO_2$, ammonium uranyl carbonate (AUC), ammonium di-uranate (ADU), and uranyl nitrate(UN) with tributylphosphate(TBP) and dodecane, which are generated by dismantling the contaminated system components and equipment of a retired uranium conversion plant in Korea Atomic Energy Research Institute (KAERI). Electrochemical decontamination for metallic wastes contaminated with uranium compounds was evaluated through the experiments on the electrolytic dissolution of stainless steel as the material of the system components in neutral salt electrolytes. The effects of type of neutral salt as the electrolyte, current density, and concentration of electrolyte on the dissolution of the materials were evaluated. Decontamination performance tests using the specimens taken from a uranium conversion plant were quite successful with the application electrochemical decontamination conditions obtained through the basic studies on the electrolytic dissolution of structural material of the system components.

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A Study on the Waste Treatment from a Nuclear Fuel Powder Conversion Plant (핵연료 분말제조 공정에서 발생하는 폐액의 처리에 관한 연구)

  • Jeong, Kyung-Chai;Kim, Tae-Joon;Choi, Jong-Hyun;Park, Jin-Ho;Hwang, Seong-Tae
    • Applied Chemistry for Engineering
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    • v.7 no.6
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    • pp.1164-1173
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    • 1996

  • Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for $CO_2$ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.

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