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The radioactivity analysis of 14C and 3H in graphite from the dismantled korea research reactor and its dose estimation

Published online by Cambridge University Press:  06 June 2009

J. Barescut ,
H. R. Kim ,
S. B. Hong ,
W. Lee ,
G. H. Chung ,
G. S. Choi  and
C. W. Lee
H. R. Kim
Affiliation:
Korea Atomic Energy Research Institute, 305-600 Daejeon, Korea
S. B. Hong
Affiliation:
Korea Atomic Energy Research Institute, 305-600 Daejeon, Korea
W. Lee
Affiliation:
Korea Atomic Energy Research Institute, 305-600 Daejeon, Korea
G. H. Chung
Affiliation:
Korea Atomic Energy Research Institute, 305-600 Daejeon, Korea
G. S. Choi
Affiliation:
Korea Atomic Energy Research Institute, 305-600 Daejeon, Korea
C. W. Lee
Affiliation:
Korea Atomic Energy Research Institute, 305-600 Daejeon, Korea
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Abstract

A high temperature combustion method was used to analyze the 14C and 3H activities in graphite and the dose assessment was carried out to determine the clearance in the conservative way. By this method, the 3H and 14C were simultaneously trapped in the nitric acid and carbosorb, respectively. Accordingly, the sample preparation time for the measurement was reduced to the half. The combustion temperature was more than 800 degrees in centigrade for obtaining total tritium and 14C in the sample. The combustion ratio was about 99% on the graphite sample with the weight of 0.1 g. Minimum detectable activity was 0.05 Bq/g for the 14C and 0.15 Bq/g for the 3H at the same background counting time. The recoveries from the combustion furnace were around 100% and 90% in 14C and 3H, respectively. The radioactivity were 2,530 ~ 3,160 Bq/g in 14C and 1,700 ~2,040 Bq/g in 3H at this experiment. The experimental uncertainty was less than 6% in both radionuclides where the furnace recovery was dominant factor. An individual effective dose from beta and gamma radionuclides was estimated by consideration of the scenario of inhalation, ingestion and external exposure. 60Co, the radioactivity of which was measured by using HPGe detector, had a predominant effect in estimating the effective dose. The estimation showed that the graphite wastes from the dismantled research reactor should be disposed of as a low level radioactive waste rather than clearance.

Type
Research Article
Information
Radioprotection , Volume 44 , Issue 5: ECORAD 2008 - Radioecology and Environmental Radioactivity , 2009 , pp. 59 - 63
Copyright
© EDP Sciences, 2009

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References

I. Croudace and Jung.-Suk. Oh, Analysis of total tritium and 14C in solid samples, GAU/RC/2022 (2005).
Hee Reyoung Kim, et al., “The Radioactivity Analysis of 14C in the Graphite Samples from the Dismantled KRR-1&2 Sites by a high temperature furnace and a LSC”, Proceedings of the 11th International Conference on Environmental Remediation and Radioactive Waste Management ICEM07, September 2-6, 2007.
MEST Notice 2001-30 (2001).
IAEA, Measurement of Radionuclides in Food and the Environment, Technical Report Series No. 295, Vienna, 1989.
Robert J. Moffat, “Describing the Uncertainties in Experimental Results”, Experimental Thermal and Fluid Science, Vol. 1, pp. 3-17, 1988.
James D. Jenkins, Measurement Uncertainty Analysis Fundamentals, Quametec Corporation, 2004.
Hugh W. Coleman and W. Glenn Steele, Experimentation and Uncertainty Analysis for Engineers, John Willey and Sons, 1989.
UNSCEAR, Sources and Effects of Ionizing Radiation, Annex A: Dose Assessment Methodologies, UNSCAER 2000, United Nation Scientific Committee on the Effect of Atomic Radiation, (2000).
IAEA Safety Series No. 111-P-1.1, Application of Exemption Principles to the Recycle and Reuse of Materials from Nuclear Facilities, IAEA, 1992.
IAEA Safety Series No. 115, International Basic Safety standards for Protection Against Ionizing Radiation and for the Safety of radiation Sources, IAEA, 1996.