• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.277-286
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• 1989

Destructive methods are used for the turnup determination of an irradiated PWR fuel. One of the methods includes U, Pu, Nd-148 and Nd-(145+146) determination by an isotope dilution mass spectrometry using triple spikes (U-233, Pu-242 and Nd-150). The method involves two sequential ion exchange resin separation procedures. Pu is eluted from the first anion exchange resin column (Dowex AG 1$\times$8) with 12 M HCl-0.1 M HI mixed solution, followed by U elution with 0.1 M HCl. Nd is isolated from other fission products on the second anion exchange resin column (Dowex AG 1$\times$4) with a nitric acid-methanol eluent. Each fraction is analysed by thermal ionization mass spectrometry. The difference between Nd-148 and Nd-(145+146) method is found with an average 2.07%. The results are compared with those by the heavy element method using U and Pu isotopes and by the destructive y-spectrometric measurement of Cs-137. The dependences of isotope composition of U and Pu on burn-up, and correlation between those isotopes are illustrated graphically.

• Journal of the Korean Vacuum Society
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• v.21 no.5
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• pp.225-232
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• 2012

It is described that the proton beam induceds micro-size defects and electronic deep levels in n or p type single crystal silicon. Positron lifetime and Coincidence Doppler Broadening Positron Annihilation Spectroscopy were applied to study of characteristics of p type and n type silicon samples. In this investigation the numerical analysis of the Doppler spectra was employed to the determination of the shape parameter, S, defined as the ratio between the amount of counts in a central portion of the spectrum and the total counts of whole spectrum. The samples were exposed by 3.98 MeV proton beams ranging between 0 to ${\sim}10^{14}$ particles. The S-parameter values strongly depend on the irradiated proton beam, that indicated the defects generate more. Positron lifetime shows that positrons trapped in vacancies and lifetime ${\tau}_2$ increased according to proton irradiation.

• Journal of Radiation Protection and Research
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• v.22 no.3
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• pp.143-151
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• 1997

As the pulse heights from a NaI(Tl) detector vary with the temperature of the measuring environment a significant change in temperature may affect the energy calibration of the spectrometer. The auto-adjustment of the channel corresponding to a pulse heights can be achieved by introducing an external reference source to compensate the temperature dependency of pulse heights, but unfavorable increases of the Compton continuum are caused due to the external source. In this study, the total absorption peaks dominant in the typical environmental gamma spectrum-239 keV from $^{212}Pb$, 351 keV from $^{214}Pb$, 1460 keV from $^{40}K$ and 2614 keV from $^{208}Tl$ for examples - were used as reference in the correction of energy calibration. With these peaks, the program to calibrate the energy of the s spectrum was developed using Microsoft Visual Basic language. The program developed here was applied to the environmental spectra measured at intervals of 30 minutes in the temperature range of from $-20^{\circ}C$ to $10^{\circ}C$ to demonstrate the validity and applicability. As a result of the test, the correction scheme appeared to be effective in the temperature changes encountered in the usual environment.

• Journal of the Korean Society of Radiology
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• v.6 no.1
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• pp.79-82
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• 2012

Iodine is one of important nuclides to be checked for radiation exposure after nuclear power facility accidents. After Chernobyl accident, it was observed that there is a greater amount of organic iodine in the atmosphere than inorganic iodine. In this study, we not only varied the amount of sample being exposed to $^{131}I$ and the duration of exposure to $^{131}I$ but also diluted the sample in distilled water and mixed the sample in kelp and liquid $^{131}I$ to measure and analyze the radiation detection levels. We concluded that the radiation levels were not high enough to be harmful to human body. The radiation from $^{131}I$ decreased over time, and we calculated the half life at 7-9 days. We found that the radiation from any sample containing $^{131}I$ was halved by up to 7days.

• Analytical Science and Technology
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• v.16 no.5
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• pp.418-425
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• 2003

Inorganic elemental content of bottom ash generated from municipal solid waste incinerator was determined by neutron activation analysis. Bottom ash samples were collected monthly from incinerator located in city D, strained out by the 5 mm sieve, dried by oven and pulverized by agate mortar. The samples were irradiated at NAA #1 irradiation hole in HANARO research reactor of Korea Atomic Energy Research Institute and the irradiated samples were measured by HPGe-gamma-ray spectrometer. From the activity of measured nuclides, 33 elements including As, Cr, Cu, Fe, Mn, Sb and Zn were determined applying activity creation formula and nuclear data. The quality control was conducted by simultaneous analysis with NIST standard reference materials.

• Journal of the Korean Chemical Society
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• v.21 no.6
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• pp.427-433
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• 1977

A neutron activation method has been developed for the simultaneous determination of chromium, iron, lanthanum, scandium and zinc in river-water samples. The sample is sealed in the silica ampoule without pretreatment and irradiated for a week at a thermal neutron flux of $1{\times}10^{13}n{\cdot}cm^{-2}{\cdot}sec^{-1}$. After cooling for about two days, the elements in the sample are sequentially extracted at different pH by 0.1M oxine-chloroform solution. The organic layers are checked by Gamma-ray spectrometry with $″3\;{\times}\;3″$ NaI (T1) detector connected to a 800-channel pulse hight analyzer. The ppb concentration of the elements in most of river-water samples could be determined by this method. The tracer study for the quantitative separation of the elements was also carried out.

• Journal of Soil and Groundwater Environment
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• v.14 no.2
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• pp.10-16
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• 2009

Characteristics of natural radioactivity were investigated for soils collected from seven sites of different bed rocks distributed in the Keum River area of Korea by the use of a Gamma-ray spectrometry. Specific activity (SA) and SA ratio (SAR) of typical naturally occurring radioactive nuclide such as $^{226}$Ra, $^{228}$Ac and $^{40}$K were determined for the soil samples. The SA values of $^{226}$Ra, $^{228}$Ac and $^{40}$K in 41 soils of 7 sites are 26.7-485 (74.2 ${\pm}$ 72.2), 30.9-157 (90.7 ${\pm}$ 32.7) and 203-1558 (990 ${\pm}$ 203) Bq/kg, respectively. The SA of $^{226}$Ra has very different values by the soils and the sites. Especially the SA of $^{226}$Ra in a soil sample of Ogcheon site is 485 Bq/kg while most SA of 41 soil samples are < 100 Bq/kg. SA of $^{228}$Ac has a little different values with the soils and sites, however the SA of $^{40}$K has almost constant values in all soil samples. The SAR values of $^{26}$Ra/$^{228}$Ac, $^{226}$Ra/$^{40}$K and $^{228}$Ac/$^{40}$K in 41 soils of 7 sites are 0.343-6.11 (0.865 ${\pm}$ 0.883), 0.0258-0.759 (0.0814 ${\pm}$ 0.1l17) and 0.0373-0.178 (0.0945 ${\pm}$ 0.0373), respectively. The SARs of $^{226}$Ra/$^{228}$Ac and $^{226}$Ra/$^{40}$K have very different values by the soils and the sites, however the SAR of $^{228}$Ac/$^{40}$K has a little difference by the soil and sites.

• Journal of Radiation Protection and Research
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• v.41 no.1
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• pp.71-79
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• 2016

Background: It is necessary to analyze radioactivity of naturally occurring radioactive materials (NORM) in products to ensure radiological safety required by Natural Radiation Safety Management Act. The pretreatments for the existing analysis methods require high technology and time. Such destructive pretreatments including grinding and dissolution of samples make impossible to reuse products. We developed a rapid and simple procedure of radioactivity analysis for thorium series in the products containing NORM. Materials and Methods: The developed method requires non-destructive or minimized pretreatment. Radioactivity of the product without pretreatment is initially measured using gamma spectroscopy and then the measured radioactivity is adjusted by considering material composition, mass density, and geometrical shape of the product. The radioactivity adjustment can be made using scaling factors, which is derived by radiation transport Monte Carlo simulation. Necklace, bracelet, male health care product, and tile for health mat were selected as representative products for this study. The products are commonly used by the public and directly contacted with human body and thus resulting in high radiation exposure to the user. Results and Discussion: The scaling factors were derived using MCNPX code and the values ranged from 0.31 to 0.47. If radioactivity of the products is measured without pretreatment, the thorium series may be overestimated by up to 2.8 times. If scaling factors are applied, the difference in radioactivity estimates are reduced to 3-24%. Conclusion : The developed procedure in this study can be used for other products with various materials and shapes and thus ensuring radiological safety.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.237-247
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• 1993

Activity ratio of two radioactive primary fission products which had sufficiently different half-lives was expressed as functions of cooling time and irradiation histories in which average burnup, irradiation time, cycle interval time and the dominant fissile material of the spent fuel were included. The gamma-ray spectra of 36 samples from 6 spent PWR fuel assemblies irradiated in Kori unit-1 reactor were obtained by a spectrometric system equipped with a high purity germanium gamma-ray detector. Activity ratio $^{l44}$Ce $^{l37}$Cs, analyzed from each spectrum, was used for the calculation of cooling time. The results show that the radioactive fission products $^{l44}$Ce and $^{l37}$Cs are considered as useful monitors for cooling time determination because the estimated cooling time by detection of activity ratio $^{l44}$Ce $^{l37}$Cs agreed well with the operator declared cooling time within relative difference of $\pm$5 % despite the low counting rate of the gamma-ray of $^{l44}$Ce (about 10$^{-3}$ count per second). For the samples with several different irradiation histories, the determined cooling time by modeled irradiation history showed good agreement with that by known irradiation history within time difference of $\pm$0.5 year. From this result, it would be expected to be possible to estimate reliably the cooling time of spent nuclear fuel without the exact information about irradiation history. The feasibility study on identification of and/or sorting out spent nuclear fuel by applying the technique for cooling time determination was also performed and the result shows that the detection of activity ratio $^{l44}$Ce $^{l37}$Cs by gamma-ray spectrometry would be usefully applicable to certify spent nuclear fuel for the purpose of safeguards and management in a facility in which the samples dismantled or cut from spent fuel assemblies are treated, such as the post irradiation examination facility.mination facility.

• The Korean Journal of Nuclear Medicine Technology
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• v.22 no.1
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• pp.28-34
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• 2018

Purpose With regard to the use of radioiodine in domestic medical institution, the case of exceeding the allowance of nuclear safety Act about radioactive concentration in drainage was found. Through understanding the cause of exceeding case and analyzing radioactive concentration in drainage, evaluating the relationship of the public waters in surroundings and usefulness. Materials and Methods From November 1, 2014 to April 30th, 2015, the research is aiming at domestic twenty hospitals for six months. By using a HPGe gamma-ray spectrometer(Canberra DSA-1000) and GENIE-2000 Analysis software for comparative analysis, measuring a radioactive concentration of radioiodine in drainage. Consequently, we confirm the excess of radioactive concentration of radioiodine in seven medical institutions. Results Conducting a survey of twenty hospitals and average radioactive concentration of radioiodine in drainage appears $42,100Bq/m^3$. The features of domestic hospitals where show a high radioactive concentration are a number of medical treatment patient when using radioactive iodine and the absence of private rest room. During I-131 whole body scan, the pretreatment procedure of urinating is considered emission of residual Iodine. In public waters, the cause of exceeding detect on radioactive concentration in drainage suppose a diagnostic radioactive iodine. Conclusion We confirm the importance of enhanced education, providing a safety control instructions and installing a private rest rooms for patients who injected a low capacity radioiodine. Also, constructing institutional and legal management system is considered about the Emission management standard in drainage.