• 방사성폐기물학회지
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• 제4권4호
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• pp.311-320
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• 2006

[ $^{14}C$ ] 핵종이 함유된 IRN-150 혼상 폐수지로부터 $H^{14}CO_3$ 이온의 제거 및 제거된 $^{14}C$ 핵종의 $^{14}CO_2$ 기체로의 전환 특성을 고찰하였다. 비방사성 $HCO_3$ 이온이 흡착되어 있는 IRN-150 혼합수지로부터 $HCO_3$ 이온의 탈착용액내로의 분리 및 $CO_2$ 기체로의 전환 특성을 용액의 농도 변화에 따라 평가하였으며, 탈착용액으로는 $NaNO_3,\;Na_3PO_4,\;NH_4H_2PO_4,\;H_3PO_4$를 사용하였고, 비교 평가를 위하여 NaOH, $HNO_3$, HCl를 이용한 $CO_2$기체로의 전환 특성을 분석하였다. 아울러 월성 원자력발전소에 저장중인 실제 폐수지를 이용하여 $NH_4H_2PO_4,\;H_3PO_4$ 탈착용액을 이용한 폐수지내 $^{14}C$ 핵종의 $^{14}CO_2$ 기체화 특성을 평가하였고, 탈착후 잔류용액내 존재하는 $^{137}Cs,\;^{60}Co$ 감마핵종을 분석하였다.

• 방사성폐기물학회지
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• 제4권4호
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• pp.373-384
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• 2006

중수로 원전내 여러 계통으로 부터 발생된 폐수지내에는 $^{14}C$ 핵종이 다량 함유되어 있으며, Class A 및 C 폐기물로 분류되는 폐수지의 적정 처리 기술 개발을 위한 기초연구를 수행하였다. IRN-150 혼상 이온교환수지를 이용하여 비방사성 $HCO_3$ 이온과 양이온의 흡착 특성 및 탈차용액을 이용한 $HCO_3$ 이온의 제거 특성을 고찰하였다. IRN-150 수지의 $HCO_3$ 이온의 흡착능은 이론값에 근접한 11 mg-C/g-IRN-150을 나타내었고, $CS^+,\;CO_2^+,\;Na^+,\;NH_4^+$ 양이온의 흡착 친화도를 단일성분 및 복합성분 시스템을 이용하여 분석하였다. 여러 가지 탈착용액을 이용한 폐수지로부터 $HCO_3$ 이온의 제거 특성을 평가한 결과, $^{14}C$ 핵종을 전량 효과적으로 제거하기 위해서는 $NaNO_3,\;Na_3PO_3$ 보다도 $NH_4H_2PO_4$ 용액이 유리한 것으로 나타났다.

• 대한핵의학회지
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• 제13권1_2호
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• pp.7-14
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• 1979

Radionuclide cardiac angiography has distinct advantages in safety, patient comfort, cost and ease of performance. This method offers diagnostic accuracy equivalent to that of cardiac catheterization. By this method the qualitative and quantitative diagnosis of the cardiac shunts are available. Also for it is repeatable with ease and more physiologic, it has application in following pre- and post-operative shunt patients. We performed the radionuclide cardiac angiographies in 147 cases of heart diseases and 26 cases of normal group. 1. The detection of left-to-right shunt was possible in 22 of 24 patients, and 2 patients were not diagnosed due to small shunt amount. (Qp/Qs<1.3) In 21 patients of right-to-left shunt, all were diagnosed by radionuclide cardiac angiography. 2. With the pulmonary time-activity curve, $C_2/C_1$ ratio was calculated. In normal control group, a range of $C_2/C_1$ ratios of $21{\sim}38%$ was established with a mean value of $28.6{\pm}4.6%$. In patients with left-to-right shunts determined by catheterization data, the range of $C_2/C_1$ ratio was $33{\sim}90%$, with a mean value of $67.8{\pm}12.2%$. 3. In 8 cases of left-to-right shunt, $Q_p/Q_s$ ratios determined by radionuclide cardiac angiography were compaired with those of cardiac catheterization. The correlation coefficient was 0.907. (P<0.001) 4. Postoperative radionuclide cardiac angiographies were done in 21 cases. 3 of 13 patients with left-to-right shunts were found to have residual shunts. 8 patients with right-to-left shunts were confirmed to have no residual shunt.

• 방사성폐기물학회지
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• 제18권4호
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• pp.439-455
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• 2020

Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.847-856
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• 2014

Spent resin waste containing a high concentration of $^{14}C$ radionuclide cannot be disposed of directly. A fundamental study on selective $^{14}C$ stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchange equilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive $HCO_3{^-}$ ion, as the chemical form of $^{14}C$, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order of $NO_3{^-}$ > $HCO_3{^-}{\geq}H_2PO_4{^-}$. Thus the competitive adsorption affinity of $NO_3{^-}$ ion in binary systems appeared far higher than that of $HCO_3{^-}$ or $H_2PO_4{^-}$, and the selective desorption of $HCO_3{^-}$ from the resin was very effective. On one hand, the affinity of $Co^{2+}$ and $Cs^+$ for the resin remained relatively higher than that of other cations in the same stripping solution. Desorption of $Cs^+$ was minimized when the summation of the metal ions in the spent resin and the other cations in solution was near saturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of the simple second waste process, $NH_4H_2PO_4$ solution was preferable for the stripping of $^{14}C$ from the spent resin.

• Journal of Radiation Protection and Research
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• 제16권2호
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• pp.27-39
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• 1991

압축된 국산벤토나이트에서 Sr-85, Cs-237, Co-60 및 Am-241의 확산연구를 수행하였다. 본 실험에서는 원통형으로 압축된 벤토나이트 시료의 중앙부에서 축 방향으로 방사성핵종의 확산이동이 이루어지도록 하여 각 방사성핵종의 확산계수를 측정하였다 그리고 벤토나이트의 열처리 온도와 압축밀도가 확산에 미치는 영향 등을 분석하였다. Sr-85, Cs-137, Co-60 및 Am-241의 겉보기 확산계수는 각각 $1.07{\times}10^{-11},\;6.705{\times}10^{-13},\;1.226{\times}10^{-13},\;1.310{\times}10^{-14},\;m^2/sec$로 측정되었다. 그리고 시료의 압축 밀도를 $1.8g/cm^2$에서 $2.0g/cm^2$으로 증가시켰을 때, Cs-137의 확산계수는 약 1/4로 감소되어 나타났다. 반면, 열처리된 벤토나이트의 경우에는 확산계수가 크게 변하지 많았는데, 이는 $150^{\circ}C$ 이하의 온도에서는 국산 벤토나이트가 방사성핵종의 이동을 지연시킬 수 있는 화학적 방벽으로서 사용할 수 있다는 가능성을 보여준 것이라 생각된다. 그리고 음이온 Cl-36의 화산계수를 이용하여 도출한 각 방사성핵종의 공극확산계수와 표면확산계수를 측정한 겉보기확산계수와 비교해 볼 때, 전체 방사성 핵종의 확산이동에 있어서 표면확산이동이 지배적인 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.269-274
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• 2022

An artificial neural network (ANN) that identifies radionuclides from low-count gamma spectra of a NaI scintillator is proposed. The ANN was trained and tested using simulated spectra. 14 target nuclides were considered corresponding to the requisite radionuclide library of a radionuclide identification device mentioned in IEC 62327-2017. The network shows an average identification accuracy of 98.63% on the validation dataset, with the gross counts in each spectrum N_c = 100~10000 and the signal to noise ratio SNR = 0.05-1. Most of the false predictions come from nuclides with low branching ratio and/or similar decay energies. If the N_c>1000 and SNR>0.3, which is defined as the minimum identifiable condition, the averaged identification accuracy is 99.87%. Even when the source and the detector are covered with lead bricks and the response function of the detector thus varies, the ANN which was trained using non-shielding spectra still shows high accuracy as long as the minimum identifiable condition is satisfied. Among all the considered nuclides, only the identification accuracy of ²³⁵U is seriously affected by the shielding. Identification of other nuclides shows high accuracy even the shielding condition is changed, which indicates that the ANN has good generalization performance.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.214-221
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• 2005

Spent ion-exchanged resin generated from various purification systems in CANDU reactor is causing concern due to a limited storage capacity and safe disposal. As a suggestion for a proper treatment technology for the spent ion-exchanged resin containing a high activity of C­14 radionuclide which would be classified as Class A and C wastes, a fundamental study for the development of C-14 removal technology from a spent resin was performed. The adsorption characteristics of the inactive $HCO_3^-$ ion and other ions in a stripping solution on IRN-150 mixed resin was evaluated and the removal technology of the $HCO_3^-$ ion adsorbed on IRN-150 by an alkaline stripping method was proposed.

• 방사성폐기물학회지
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• 제20권4호
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• pp.489-500
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• 2022

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly ³H (500-3,600 Bq·g^-1), ¹⁴C (7.5-29 Bq·g^-1), ⁵⁵Fe (1.1- 7.1 Bq·g^-1), ⁵⁹Ni (0.60-1.0 Bq·g^-1), ⁶⁰Co (0.74-70 Bq·g^-1), ⁶³Ni (0.60-94 Bq·g^-1), ⁹⁰Sr (0.25-5.0 Bq·g^-1), ¹³⁷Cs (0.64-8.7 Bq·g^-1), and ¹⁵²Eu (0.19-2.9) Bq·g^-1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32-1.1 Bq·g^-1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

• 핵의학기술
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• 제16권1호
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• pp.102-107
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• 2012

단 반감기 핵종을 이용한 PET검사는 방사성동위원소의 빠른 물리적 붕괴로 인하여 영상 획득을 위한 계수검출이 제한적이다. 이러한 이유로 비교적 낮은 감도의 검사에서는 보다 정확한 정량적 평가를 위하여 긴 시간동안 영상 획득을 적용하기도 한다. 본 연구에서는 $^{11}C$와 $^{18}F$를 이용한 PET 검사 시 영상 획득 시간에 따른 차이를 평가하여 합리적인 영상 획득 시간에 관하여 알아보고자 한다. 1994 NEMA Phantom에 $^{11}C$은 $30.08{\pm}4.22MBq$, $^{18}F$은 $40.08{\pm}8.29MBq$을 증류수에 희석하여 채운 후 $^{11}C$은 동적영상 1분씩 20회, 정적 영상 20분, $^{18}F$은 동적영상 2분30초씩 20회, 정적영상 50분을 획득하였다. 모든 데이터는 동일한 재구성법을 적용하였으며, 시간의 경과에 따른 붕괴보정을 적용하였다. 방출영상에 관심영역을 설정하고 최대 방사능 농도값(kBq/mL)을 비교하였으며, 각각의 동적영상을 영상 획득 시간의 증가에 따라 1개씩 증가시켜 영상 합산(Image summation) 후 영상의 관심 영역 내에서의 최대 방사능 농도값(kBq/mL)을 평가하였다. $^{11}C$ 동적영상의 시간 경과에 따른 최대 방사능 농도값은 $3.85{\pm}0.45{\sim}5.15{\pm}0.50kBq/mL$, 정적영상은 $2.15{\pm}0.26kBq/mL$였다. $^{18}F$ 동적영상은 $9.09{\pm}0.42{\sim}9.48{\pm}0.31kBq/mL$, 정적영상은 $7.24{\pm}0.14kBq/mL$였다. $^{11}C$의 동적영상 합산에서 영상 획득 시간의 합이 5, 10, 15, 20분으로 증가할수록 $2.47{\pm}0.4$, $2.22{\pm}0.37$, $2.08{\pm}0.42$, $1.95{\pm}0.55kBq/mL$으로 감소하였으며, $^{18}F$의 경우 합산된 영상 획득 시간의 합이 12분 30초, 25분, 37분 30초, 50분으로 증가할수록 $7.89{\pm}0.27$, $7.61{\pm}0.23$, $7.36{\pm}0.21$, $7.31{\pm}0.23kBq/mL$으로 감소하였다. 영상의 질을 평가 하는 SNR에서는 $^{11}C$과 $^{18}F$ 모두 동적영상획득 방법에서는 주사 후 시간이 흐를수록 SNR가 저하 되었으나, 영상 합산획득 방법에서는 합산 횟수가 증가 할수록 SNR가 향상 되는 것을 알 수 있었다. 동적영상에서 시간 경과에 따른 최대 방사능 농도값은 $^{11}C$과 $^{18}F$에서 증가하였고, 동적영상 합산의 경우는 합산수가 증가함에 따라 최대 방사능 농도값은 $^{11}C$과 $^{18}F$ 감소함을 보였다. $^{18}F$을 이용할 경우에는 시간 경과에 따른 정량평가의 오차를 크게 고려하지 않아도 될 것으로 사료되고, $^{11}C$를 이용한 PET 검사는 시간경과에 따른 감쇠 보정의 오차를 감안하여 추가의 감쇠 보정법을 적용하거나 30%정도의 오차를 적용하여 정적영상 획득시간을 반감기의 25% 이내인 5분 내외로 설정해야 할 것이다.