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.

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.

Implant-supported overdenture with prefabricated bar attachment system in edentulous mandibular patient

Seung-Ryong Ha,Sung-Hun Kim,Seung-Il Song,Seong-Tae Hong,Gy-Young Kim 대한치과보철학회 2012 The Journal of Advanced Prosthodontics Vol.4 No.4

Implant-supported overdenture is a reliable treatment option for the patients with edentulous mandible when they have difficulty in using complete dentures. Several options have been used for implant-supported overdenture attachments. Among these, bar attachment system has greater retention and better maintainability than others. SFI-Bar(R) is prefabricated and can be adjustable at chairside. Therefore, laboratory procedures such as soldering and welding are unnecessary, which leads to fewer errors and lower costs. A 67-year-old female patient presented, complaining of mobility of lower anterior teeth with old denture. She had been wearing complete denture in the maxilla and removable partial denture in the mandible with severe bone loss. After extracting the teeth, two implants were placed in front of mental foramen, and SFI-Bar(R) was connected. A tube bar was seated to two adapters through large ball joints and fixation screws, connecting each implant. The length of the tube bar was adjusted according to inter-implant distance. Then, a female part was attached to the bar beneath the new denture. This clinical report describes two-implant-supported overdenture using the SFI-Bar(R) system in a mandibular edentulous patient.

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.

Preliminary Conceptual Design of a Laboratory for Measuring the Ultra Low-Level Radioactivity

Seung-gy Ro,In-suk Suh,Jae-hyung Yoo,Kun-joong Yoo,Eun-ka Kim,Sung-soo Cho 한국방사성폐기물학회 2014 한국방사성폐기물학회 학술논문요약집 Vol.12 No.1

Determination of the Spontaneous Fission Rate of 238U Using Solid State Track Recorder

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

Preliminary Conceptual Design of a Laboratory for Measuring the Ultra Low-Level Radioactivity

Seung-gy Ro,In-suk Suh,Jae-hyung Yoo,Kun-joong Yoo,Eun-ka Kim,Sung-soo Cho 한국방사성폐기물학회 2014 한국방사성폐기물학회 학술논문요약집 Vol.12 No.2

Prompt Fission Neutron Spectra in Supercritical Accidents (Influence on the Fission Spectrum-averaged cross-sections of Some Threshold Activation Reactions)

Ro, Seung-Gy,Jun, Jae-Shik Korean Nuclear Society 1975 Nuclear Engineering and Technology Vol.7 No.2

핵임계 사고시에 방출되는 즉발중성자 스펙트럼을 두개의 스펙트럼 매개변수를 갖는 일반화된 Cranberg 식으로 표시할 수 있다고 가정한 다음, 이들 매개변수를 변화시키면서 몇개의 발단 방사화검출체에 대한 평균핵반응단면적의 변화를 고속전자계산기로 계산하였다. 평균핵반응 단면적은 스펙트럼 변화에 따라 민감하게 변화하는데 발단 방사화 에너지가 높을수록 그 변화정도가 심한것 같았다. On the assumption that the spectral distribution of prompt fission neutrons released from supercritical accidents can be expressed by the generalized Cranberg form with two spectral parameters, which is then transformed into the single parameter form, a variation of the fission spectrum-averaged cross-sections for some threshold reactions with varying the spectral parameter has teen calculated using an electronic computer. It appears that the average cross-sections are very sensitive to the spectral deformation, especially those for the detectors having the threshold at high neutron energy are high compared to those for the detectors of which the threshold energies are comparatively low.

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.

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.