수중 방사능 측정시 이온교환농축법과 증발건조법의 비교

지평국,박종묵,노성기 대한방사선 방어학회 1988 방사선방어학회지 Vol.13 No.2

수중의 방사능을 측정하기 위한 전처리과정으로서 이온교환농축법과 증발건조법을 서로 비교하였다. 시료를 증발건조법으로 처리하였을 때 방사성물질의 손실율은 이온교환농축법에 비해 20% 정도 많았다. 또, 1리터의 시료를 처리하는데 소요되는 증발시간은 70℃에서 증발시킨 경우 약20시간이었으나 이온교환농축법으로 같은 양의 시료를 처리하는데 소요된 시간은 약6시간이었다. 따라서 이온교환농축법이 증발건조법에 의해 효과적이며 특히 수중의 저준위 방사성물질 측정에 적합한 것으로 나타났다. An ion-exchange method for the detection of radioactivity in water using ion-exchange resion in concentrating radioactive nuclides was compared with an evaporation method. The loss of the radioactive materials in the sample treated by the ion-exchange method was less by about 20% than that by the evaporation method. In addition, the evaporation method needed about 20 hours for evaporating one liter of the sample at 70℃, while the ion-exchange method spent 6 hours to adsorb and desorbs the same amount of the sample on the resion. Consequently, the ion-exchange method is more effective than the evaporation method for the treatment of the radioactively contaminated water and is especially suitable for detecting the low-level radioactivity in water.

0.412 MeV 감마선에 대한 원주형 NaI(Tl) 섬광체의 총 절대검출효율 계산

홍권표,신희성,이상윤,노성기 대한방사선 방어학회 2002 방사선방어학회지 Vol.27 No.4

Total absolute detection efficiencies of a 7.62 cm(dia.) and 7.62 cm(height) cylindrical NaI(Tl) crystal have been calculated for 0.412 MeV r -rays from a source(point, circular disk, square and line type). In this calculation the linear energy-absorption coefficients based on Hubbell's data have been considered and then calculated total absolute detection efficiencies compared with those from Grosjean and Bossaert. Besides, the source axis-to-detector axis shift distance which, could give rise to about 0.05% deviation in the total absolute detection efficiencies has been calculated for a line-type source of 0.5 cm in its length when a source-to-detector distance is 5 cm. It is revealed that the total absolute detection efficiencies obtained in this study are considerably different from those of Grosjean and Bossaert. In addition it is found that the deviation induced due to an imperfect center of a line type source may be within 0.05% if the shifted discrepancy is no larger than 1.74mm.

Fast Neutron Dosimetry with Two Threshold Detectors in Criticality Accidents of Nuclear Reactors

Ro, Seung-Gy Korean Nuclear Society 1970 Nuclear Engineering and Technology Vol.2 No.2

An attempt has been made to do interpretation of the fast neutron dose with two threshold detectors incorporated with the Harwell criticality locket. This method is based on the assumption that the spectral distribution of fission neutrons in criticality accidents may be governed by one spectral parameter. The surface-absorbed dose for a unit fission neutron fluence seems to be insensitive to spectral shifts of the fission neutron spectrum. The average cross-sections for the activation detectors, however, are considerably changed with the neutron spectral shape, which may lead to a large error in calculating the dose from the reaction rate if one uses a fixed value for the average cross sections regardless of the neutron spectral distribution. Besides, the doses calculated from three representative formulae for fission neutron spectra have been compared : these formulae are Watt, Cranberg at al. and Maxwellian forms. The results obtained front the Maxwellian formula show a departure from the Watt and Cranberg's, both being similarly close.

Average and Effective Energies, and Fluence-Dose Equivalent Conversion Factors for $^{239}Pu-Be,\;^{241}Am-Li\;and\;^{241}Am-F$ Neutron Sources

Ro, Seung-Gy,Yoo, Young-Soo Korean Nuclear Society 1971 Nuclear Engineering and Technology Vol.3 No.3

Average and effective energies for 239Pu-Be, 241Am-Li and 241Am-F neutron sources have been calculated from a number of published data for the neutron spectra and for the dose equivalent as a function of neutron energies by a numerical method. Also a calculation of the dose equivalent conversion factors, i. e., the first collision dose equivalent and the surface (or multicollision) dose equivalent that equals the product of surface-absorbed dose and a corresponding quality factor, per unit fluence of neutrons from these sources has been carried out in the same way as before. The results are as follows : 1. for average energies 4.07$\pm$0.33, 0.42 and 1.41 MeV; 2. for effective energies based on the concept of the first collision process in the human body 4.45$\pm$0.344, 0.51 and 1.47 MeV; 3. for effective energies based on the concept of the multi-collision process in the human body 4.50$\pm$0.36, 0.50 and 1.45 MeV; 4. for fluence-first collision dose equivalent conversion factors (2.74$\pm$0.07)10$^{-8}$ , 1.58$\times$ 10$^{-8}$ and 2.34$\times$10$^{-8}$ rems/(n/$\textrm{cm}^2$); and 5. for fluence-surface dose equivalent conversion factors (3.55$\pm$0.09)10$^{-8}$ , 2.19$\times$10$^{-8}$ and 2.82$\times$10$^{-8}$ rems/(n/$\textrm{cm}^2$) : respectively.

A Method for Determining Dead Times of a G.M. Defector as a Function of the Count Rate

Ro, Seung-Gy Korean Nuclear Society 1971 Nuclear Engineering and Technology Vol.3 No.1

A method for determining dead times of a G.M. detector as a function of the count rate has been investigated using the Mn$^{56}$ radioactive sample. The formula, (equation omitted), seems to be useful for determining a relation between the dead time and the count rate. Here (equation omitted)(N$_1$) is the dead time for the count rate N$_1$, N$_1$is the count rate at time zero, Nt is the count rate at time t, λ is the radioactive decay constant of the sample used, and t is the time between the first and second runs. When all the counting data were corrected for the dead times evaluated with this formula and then a variation of these corrected counting data with rime was observed, the results showed quite a good agreement with the published data for the radioactive decay of Mn$^{56}$ . Besides, it appears that the dead time decreases as the count rate increases in a dead time-to-count rate relation obtained by the same formula.

Fast Neutron Dosimetry in Criticality Accidents

Ro,Seung Gy,Yook,Chong Chul 대한방사선방어학회 1976 방사선방어학회지 Vol.1 No.1

核臨界事故時에 放出되는 速中性子가 散亂中性子로 重??되어 있는 狀態에서 放射化 및 發端放射化檢出器를 利用하여 速中性子를 測定 및 解析할 수 있는 한 方法을 提案하였으며 이 測定에 있어서 主要因子, 卽 몇 개의 發端放射化檢出器에 對한 平均核反應斷面積과 中性子當線量換算係數를 電子計算機로 計算하였다. 그 結果 核分裂中性子의 스펙트럼 測定에는 發端에너지가 높은 檢出器가 有利한 것에 反해 發端에너지가 낮은 것은 散亂媒質이 없는 核臨界裝置의 事故時에 있어서 速中性子의 時積分線束密度 測定界로서 有用한것 같았다. 그리고 硫黃의 (n,p) 核反應에 對한 平均斷面積은 核分裂 中性子의 解析的 表現式에 無關한 것처럼 보였다. 그밖에 中性子當 線量換算係數의 變化는 核分裂 中性子 스펙트럼의 解析的 表現式과 核分裂形態에 따라 敏感하게 變化되지 않은 것 같았다. A suggestion has been made for neutron dosimetric techniques using activation and threshold detectors in criticality accidents. Neutron dosimetrical parameters, namely, the fission spectrum-averaged cross-sections of some threshold reactions and fluence-to-dose conversion factors have been calculated by the use of an electronic computer. It appears that detectors having comparatively high threshold energy give more fine information on spectral deformation in criticality accidents, while detectors with low threshold energy are of usefulness for measuring fast neutron fluence regardless of fissioning types. Unexpectedly it is found that the fission spectrum-averaged cross sections of the 32S(n,p)32P reaction is not sensitive to analytical forms of fission neutron spectrum: the modified Cranberg and Maxwellian forms. In addition, the fluence-to-dose conversion factors seem to be insensitive to both spectral functions and fissioning types.

Atom Number Densities for Uranyl Nitrate Solution

Seung Gy Ro,Duck Kee Min,Jung-Kyoon Chon Korean Nuclear Society 1982 Nuclear Engineering and Technology Vol.14 No.3

여러가지 질산우라늄용액에 대한 우라늄의 용존농도, 질산의 노르말농도 및 용액의 밀도등을 측정하여 얻은 결과를 최소자승법으로 분석한 후 우라늄의 용존농도와 질산의 노르말농도만을 알므로서 질산우라늄용액속에 들어있는 물의 함량을 결정할 수 있는 실험식, Q=1-0.3628C-0.0327H$^{+}$,을 유도하였다. 여기서 Q, C 및 H$^{+}$는 각각 물함량(g/cc), 우라늄의 용존농도(g/cc)및 질산의 노르말농도를 뜻한다. 그리고 이 유도식을 써서 임의 우라늄용액에 대한 구성원소별 원자수밀도와 핵임계도를 산출하고 그 결과를 우라늄의 용존농도, 질산의 노르말농도 및 용액의 밀도를 근거로 하여 얻은 값과 비교해 보았다. 그 결과 유도식은 우라늄의 용존농도 0.004~0.2959g/cc 및 질산의 노르말농도 1.00~5.06사이에서 유용하게 쓰일 수 있을 것으로 보였다. An empirical formula for determining water content as functions of uranium concentration and nitric acid normalities in uranyl nitrate solutions has been derived from a least-squares analysis of experimental data, i.e., uranium concentration, nitric acid normalities and solution densities for a large number of UO$_2$(NO$_3$)$_2$ solutions. The formula derived is Q=1-0.3628C-0.0327H$^{+}$ where Q, C, and H$^{+}$ stand for water content (g/cc), uranium concentration (g/cc), ana nitric acid normality, respectively. Atom number densities and nuclear criticality for hypothetical uranyl nitrate solutions have been calculated by using the empirical formula, ana compared with the results obtained on the basis of uranium concentration, nitric acid normality, and solution density. The empirical formula derived in this study seems to be useful in uranium concentrations ranging from 0.295g/cc down to 0.004g/cc and nitric acid normality from 5.06 to 1.00..00.

Fabrication and Evaluation of Radiation Shielding Property of Epoxy Resin-Type Neutron Shielding Materials

Ro,Seung-Gy,Do,Jae-Bum,Yoon,Jeong-Hyoun,Choi,Byung-Il,Cho,Soo-Haeng 대한방사선 방어학회 1997 방사선방어학회지 Vol.22 No.2

사용후핵연료 수송용기등에 사용되는 에폭시수지계 중성자 차폐재, KNS(Kaeri Neutron Shield)-101, KNS-102 및 KNS-103를 제조하였다. 기본물질은 에폭시수지이며, 첨가제로는 폴리프로필렌, 수산화알루미늄 및 탄화붕소이다. 이들 중성자 차폐재들은 유동성이 좋아 수송용기와 같은 복잡한 구조에 사용할수 있다. 제조된 중성자 차폐재들을 가압경수로 사용후핵연료 28다발을 수송할 수 있는 수송용기에 적용하여 차폐능 평가를 수행하였다. 세가지 중성자 차폐재를 수송용기에 적용하여 ANISN 코드로 차폐능 평가를 수행한 결과 정상수송시 중성자 차폐재의 두계가 10cm이상 일 때 수송용기 반경방향표면에서 최대 방사선량율은 300 μSv/h로 나타났으며, 수송용기 표면에서 100cm 지점에서의 최대 방사선량율은 97 μSv/h로 나타났다. 이들은 모두 관련된 법규들에서 규정된 최대허용방사선량율을 만족하는 것으로 나타났다. Epoxy resin-type neutron shielding materials, KNS(Kaeri Neutron Shield)-101, KNS-102, and KNS-103 have been fabricated to be used in spent fuel shipping cask. The base material is epoxy resin, and polypropylene, aluminium hydroxide, and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to apply this resin shield to complicated geometric shapes such as shipping cask. The shielding property of these shielding materials for shipping cask for loading 28 PWR spent fuel assemblies has been evaluated. ANISN code is used to evaluate the shielding property of the shipping cask with the thickness of the three neutron shielding materials greater than 10 cm. As a result of analysis, the maximum calculated dose rate at the radial surface of the cask is determined to be 300 μSv/h and the maximum calculated dose rate at 100 cm from the cask is 97 μSv/h. These dose rates remain within allowable values specified in related regulations.

Determination of the Spontaneous Fission Rate of ?? Using Solid State Track Recorder

Ro,Seung Gy,Koh,Byungryung,Yook,Chong Chul 대한방사선 방어학회 1985 방사선방어학회지 Vol.10 No.2

固體飛跡檢出器인 雲母를 238UO2 箔板에 約 5年間 密着시켜 두었다가 雲母만을 꺼내어 弗酸속에서 腐蝕시킨후, 光學顯微鏡으로 그 속에 생긴 核分裂破片의 飛跡을 觀測하여 233U의 自發核分裂率을 決定하였다. 이 實驗에서 얻은 238U의 自發核分裂率은 5.21±0.33 fissions/g-sec이었다. The spontaneous fission rate of 238U has been determined using a solid state track recorder that was a pre-etched mica. Counting the tracks revealed in mica made it possible to calculate the spontaneous fission rate. The mica remained in close contact with a 238UO2 foil for about five years. The resulting fission rate was 5.21±0.33 fissions/g-sec.

Measurement of the fast Neutron Flux Density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II Reactor Using Solid State Track Detector

Ro, Seung-Gy,Jun, Jae-Shik,Cho, Sae-Hyung Korean Nuclear Society 1973 Nuclear Engineering and Technology Vol.5 No.4

The horizontal distribution of the fast neutron flux density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II reactor at the steady power of 250 KW has been measured using a solid state track detector which is natural mica placed in contact with $^{232}$ Th fissile foil. The neutron flux density was calculated on the assumption that the fast neutron spectrum is similar to that from the thermal-induced $^{235}$ U fission. The resulting flux density distribution along the horizontal line from the center of the thermalizing column door is presented in tabular and graphical forms.