• Journal of Radiation Protection and Research
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• 제6권1호
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• pp.8-24
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• 1981

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute(ANSI) N666. These data are used to calculated the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from $2.5{\times}10^{-8}$ to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoetiergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be a useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions.

• Journal of Radiation Protection and Research
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• 제34권3호
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• pp.144-150
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• 2009

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

• Nuclear Engineering and Technology
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• 제2권2호
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• pp.85-95
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• 1970

두개의 threshold detector로서 인자로의 폭발사고시에 방출되는 속 중성자의 속도분포를 측정하고 그로부터 속 중성자의 인체흡수선량을 계산하였다. 이때 속 중성자의 속도분포는 하나의 스펙트럼 매개변수에 의하여 결정된다는 가정으로부터 얻어지는데 이 매개변수는 threshold detector의 반응율을 측정하므로서 구해진다. 속 중성자의 인체흡수선량은 속 중성자의 속도분포 변화에 따라 큰 변동이 없었으나 threshold detector의 평균반응단면적은 크게 변하였다. 따라서 속 중성자의 속도분포에 관계없이 threshold detector의 평균반응단면적을 고정된 값으로 취하여 속 중성자선량을 계산한다면 큰 오차를 일으키게 될 것이라는 것을 보여주었다. 한편 핵분열에서 방출되는 속 중성자의 속도분포에 대한 세 해석적 표현인 즉 Watt, Cranberg및 Maxwellian 공식들로부터 속 중성자 선량을 계산하여 서로 비교하였다. Watt 및 Cranberg 공식들로 부터 얻어진 속 중성자선량은 Maxwellian 공식으로부터 얻어진 그것보다 약간 높은 값을 보여 주었으며 Watt 공식에 의한 선량계산치는 Cranberg 공식에 의한 그것과 비슷한 값을 보여주었다.

• Journal of Radiation Protection and Research
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• 제14권2호
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• pp.51-62
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• 1989

The high efficiency moderated-type neutron spectrometer and doseequivalent counter were developed for the measurement of low level environmental neutrons. By using these detectors, the neutron energy spectra and dose equivalent rates due to skyshine effect were measured in the environment surrounding the accelerator facilities and also the altitude variation of cosmic ray neutrons in the aircraft flying over Japan.

• 대한약학회:학술대회논문집
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• 대한약학회 2003년도 Proceedings of the Convention of the Pharmaceutical Society of Korea Vol.2-2
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• pp.109.1-109.1
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• 2003

The aim of this study was to investigate the biological effects of blood and testis by neutron or gamma-ray irradiation in black mouse. Six-week-old C57BL male mice were irradiated with neutron (flux: 1.036739E+09) or Co60 gamma rays(dose rate: lGy/min.) The irradiation method of animal was abdominal irradiation and dose of irradiation was 10 and 20 Gy added with 5 and 15Gy in neutron irradiation.. After that, the mice were sacrificed 3 days later. Blood and testis were taken and then composition of blood in blood cell were investigated. (omitted)

• 대한방사성의약품학회지
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• 제6권1호
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• pp.10-19
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• 2020

Boron Neutron Capture Therapy (BNCT) has the advantage of selectively removing cancer cells ingesting boron compounds. In this study, the benefits for treatment time and boron compound injection dose were compared between current neutron sources and a high current neutron sources to be developed in near future. The time-activity curve (TAC) of GBM (Glioblastoma) for one bolus injection was obtained by applying modified 3 compartment model. The treatment time was determined for an accelerator-based neutron sources at the present time and a high current accelerator based neutron source to be developed in the near future. In the case of the double amount of IAEA-recommended neutron flux, the treatment time was shortened to 15 minutes. In the case of high current accelerators, which are five times the amount of IAEA-recommended neutron flux, the irradiation time is within 5 minutes. The use of a high current accelerator based neutron source in BNCT is advantageous in terms of treatment time. In addition, it can increase the efficiency of use of neutrons and reduce the boron compound injection dose to patients, thus reducing pharmacological toxicity.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.211-215
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• 2016

Background: The moderated neutron calibration fields using $^{241}Am$-Be sources and a graphite moderator have been constructed at the Facility of Radiation Standard (FRS) in the Japan Atomic Energy Agency (JAEA). Materials and Methods: The neutron spectra of the fields were evaluated by the Monte-Carlo calculations and measurements using the Bonner Multi-sphere Spectrometer. Results and Discussion: The fields have continuous neutron spectra from several MeV to thermal neutron energy, with fluence-averaged energies of 0.84 MeV and 0.60 MeV. Reference values of fluence rates and ambient/personal dose equivalent rates were determined from neutron spectra by measurements. Conclusion: Currently, the fields are available for calibration or performance test of neutron measuring instruments.

• 한국재료학회지
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• 제27권4호
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• pp.216-220
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• 2017

The effect of electron beam (EB) irradiation on the electrical properties of Zn-Sn-O (ZTO) thin films fabricated using a sol-gel process was investigated. As the EB dose increased, the saturation mobility of ZTO thin film transistors (TFTs) was found to slightly decrease, and the subthreshold swing and on/off ratio degenerated. X-ray photoelectron spectroscopy analysis of the O 1s core level showed that the relative area of oxygen vacancies ($V_O$) increased from 10.35 to 12.56 % as the EB dose increased from 0 to $7.5{\times}10^{16}electrons/cm^2$. Also, spectroscopic ellipsometry analysis showed that the optical band gap varied from 3.53 to 3.96 eV with increasing EB dose. From the results of the electrical property and XPS analyses of the ZTO TFTs, it was found that the electrical characteristic of the ZTO thin films changed from semiconductor to conductor with increasing EB dose. It is thought that the electrical property change is due to the formation of defect sites like oxygen vacancies.

• 한국의학물리학회지:의학물리
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• 제18권2호
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• pp.87-92
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• 2007

최대출력 30 MW, 하나로(HANARO) 다목적 연구용 원자로의 접선 중성자공에 붕소중성자포획치료(Boron Neutron Capture Therapy, BNCT)를 위한 열중성자 조사장치가 개발되었다. BNCT 조사장치에서는 서로 다른 물리적 특성과 생물학적 효과비를 가진 여러 성분의 방사선이 방출되기 때문에 정확한 투여선량을 결정하기 위해서는 각 성분의 정량적 분석이 필수적이다. 따라서 본 연구에서는 방사화 분석, 열형광선량계 및 이온전리함 등 여러 유형의 검출기를 사용하여 BNCT 조사장치에서 방출되는 열중성자 및 감마선 혼합장의 선량 성분을 분리, 측정하였다. 선량측정은 물 속에 함유된 불순물과 중성자의 이차반응을 최소화하기 위해 증류수를 채운 물팬텀을 이용하였다. 그리고 측정 결과는 MCNP4B 전산계산의 결과와 상호 비교하였다. 측정 결과 열중성자속은 물팬텀 10 mm와 20 mm 깊이에서 각각 $1.02E9n/cm^2{\cdot}s$과 $6.07E8n/cm^2{\cdot}s$이었고, 고속중성자선량율은 10 mm 깊이에서 0.11 Gy/hr로 미세하였다. 감마선량률은 물팬텀 20 mm 깊이에서 5.10 Gy/hr로 나타났다. 측정된 중성자와 감마선량값은 MCNP의 결과와 5% 이내로 잘 일치하였고, 열중성자속은 14%의 비교오차를 나타내었다. 이러한 결과들은 중성자 검출의 난이도를 고려할 때 충분히 신뢰할 수 있는 수준이라 판단되며, BNCT 임상 연구를 위한 선량평가 자료로 활용할 수 있을 것으로 사료된다.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1563-1570
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• 2017

In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP) code. The dose rate for the dual-purpose cask utilizing the recently developed materials of $epoxy/clay/B_4C$ and $epoxy/clay/B_4C/carbon$ fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of $epoxy/clay/B_4C$ instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.