• Nuclear Engineering and Technology
• /
• 제47권7호
• /
• pp.939-944
• /
• 2015

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a $^{252}Californium$ ($^{252}Cf$) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1-9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.

• 대한화학회지
• /
• 제10권1호
• /
• pp.1-3
• /
• 1966

The organic yields (i.e. fraction of nuclear events resulting in organic compound formation) of the radioative neutron capture reactions of halogens in purified bromobenzene have been determined varying extraction time, at $100^{\circ}C$ for thermal effect, varying irradiation time, varying neutron flux and with additional U. V. irradiation. Among the important results are; (1) The organic yields show no remarkable fluctuations with time following neutron irradiation; (2) The organic yields show no change with thermal energy; (3) The organic yields of degassed samples are same in different length of irradiation time whereas the yields of the samples in open air appear to increase with increasing time of irradiation (4) The organic yields increase remarkably with increased neutron flux; (5) The organic yields show a sharp increase by additional U. V. irradiation after neutron irradiation.

• 대한방사성의약품학회지
• /
• 제8권2호
• /
• pp.113-118
• /
• 2022

Gadolinium neutron capture therapy (Gd-NCT) is a precision radiation therapy that kills cancer cells using the neutron capture reaction that occurs when ¹⁵⁷Gd hits thermal neutrons. ¹⁵⁷Gd has the highest thermal neutron capture cross section of 254,000 barns among stable isotopes in the periodic table. Another stable isotope, ¹⁵⁵Gd, also has a high thermal neutron trapping area (~ 60,700 barns), so gadolinium that exists in nature can be used as a Gd-NCT drug. Gd-NCT is a mixed kinetic energy of low-energy and high-energy ionizing particles, which can be uniformly distributed throughout the tumor tissue, thereby solving the disadvantage of heterogeneous dose distribution in tumor tissue. The Gd complexes of small-sized molecule are widely used as contrast agents for magnetic resonance imaging (MRI) in clinical practice. Therefore, these compounds can be used not only for diagnosis but also therapy when considering the concept of Gd-NCT. This multifunctional trial can look forward to new medical advance into NCT clinical practices. In this review, we introduce gadolinium compounds suitable for Gd-NCT and describe the necessity of image guided Gd-NCT.

• Composites Research
• /
• 제36권6호
• /
• pp.435-440
• /
• 2023

본 연구에서는 기존 사용후핵연료(Spent nuclear fuel) 운반/저장 용기에 사용되는 중성자 흡수용 B₄C/Al 복합소재의 열전도도를 개선하기 위해 탄화붕소(B₄C)와 입방정 질화붕소(cBN)를 동시에 강화재로 사용한 알루미늄(Al) 기지 복합소재를 제조하고 평가를 진행하였다. 이를 위해서 교반주조 공정을 통해 복합재 잉곳을 제조하고 이를 압연하여 중성자 흡수용 소재를 성공적으로 제조하였다. 제조된 소재의 평가를 위해 cBN 첨가에 따른 열전도도와 중성자 흡수능 변화를 관찰하였다. 열전도도 측정 결과, B₄C 단일 입자만을 사용한 복합소재 대비 B₄C, cBN을 함께 사용한 복합소재가 동일 체적률 조건 하에서 약 3%의 열전도도 증가가 발생하는 것을 확인하였으며 중성자 흡수 단면적 계산을 통해 중성자 흡수능이 무시할 수 있는 수준으로 저하가 발생하는 것을 확인하였다. 본 연구의 결과를 바탕으로, 중성자 흡수 소재의 새로운 설계 방안을 제시하고 고성능 운반/저장 용기의 개발에 기여할 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
• /
• 제36권3호
• /
• pp.203-209
• /
• 2004

In-phantom neutron flux distribution is measured at the HANARO BNCT irradiation facility. The measurements are performed with Au foil and wires. The thermal neutron flux and Cd ratio obtained at the HANARO BNCT facility are $1.19{\times}10^9\;n/cm^{2}s$ and 152, respectively, at 24 MW reactor power. The measured in-phantom neutron flux has a maximum value at a depth of 3 mm in the phantom and then decreases rapidly. The maximum flux is about $25\%$ larger than that of the phantom surface, and the measured value at a depth of 22 mm in the phantom is about a half of the maximum value. In addition, the neutron beam is limited well within the aperture of the neutron collimator. The two-dimensional in-phantom neutron flux distribution is determined. Significant neutron irradiation is observed within 20 mm from the phantom surface. The measured neutron flux distribution can be utilized in irradiation planning for a patient.

• 한국재료학회지
• /
• 제9권4호
• /
• pp.404-408
• /
• 1999

Effects of heating time and radiation under high temperature on the properties of thermal, mechanical and shielding ability of modified (KNS-101), hydrogenated bisphenol-A(KNS-201) type epoxy resin and phenol-novolac(KNS-301) type epoxy resin based neutron shielding materials that are used for shipping casks for radioactive material have been investigated. At early stages, the offset temperatures of KNS-101, KNS-201 and KNS-301 increased with the heating time under high temperature, but it was rarely affected by the heating time in the later stages. In addition, the thermal conductivities of KNS-101 and KNS-201 decreased with heating time, but that of KNS-301 increased with the heating time. On the contrary, the thermal expansion coefficients of neutron shielding materials decreased with heating time. At the high temperature, the tensile strength and flexural strength of the shielding materials decreased with heating time. On the contrary, the thermal expansion coefficients of neutron shielding materials decreased with heating time. At the high temperature, the tensile strength and flexural strength of the shielding materials of KNS-101 and KNS-301 increased with heating time, but those of KNS-201 decreased with heating time. The shielding ability of neutron shielding materials slightly increased with the radiation dose, and shielding abilities of shielding materials of KNS-101 and KNS-201 were affected to a more extent than that of KNS-301 by radiation dose under high temperature.

• Journal of Radiation Protection and Research
• /
• 제19권1호
• /
• pp.13-22
• /
• 1994

CANDU 6 중수형 원자로 운전중에 Calandria Shell내에서 발생하는 $(n,\;{\gamma})$ 반응유발 열중성자속분포와 CANDU 6 발전소의 측면 및 하단 차폐구조에서의 방사선 선량률을 계산하기 위하여 몬테칼로 방법을 이용한 MCNP 4.2 코드를 사용하였다. 계산결과, Mainshell, Annular Plate와 Subshell내 의 열중성자속분포는 $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$로 나타났고, 이는 DOT 4.2 코드의 계산결과와 비교해 볼 때 약간 큰 값들의 분포를 보여주고 있다. 이 계산결과의 응용으로서 작업자 접근가능지역 (Worker Accessible Areas)에서의 감마선량률을 계산해본 결과 설계목표치인 $6{\mu}Sv/h$보다 낮은 값을 주는 것으로 나타났다. $(n,\;{\gamma})$ 반응유발 열중성자속분포에 대한 MCNP 4.2 코드의 계산결과는 CANDU 6형 원자로의 방사선 차폐해석에 중요한 자료로 널리 이용될 수 있을 것이다.

• Nuclear Engineering and Technology
• /
• 제52권3호
• /
• pp.499-507
• /
• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

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

• 한국전기전자재료학회논문지
• /
• 제31권4호
• /
• pp.198-202
• /
• 2018

In this work, we have investigated the effect of a 30-min thermal anneal at $550^{\circ}C$ on the electrical characteristics of neutron-irradiated 4H-SiC MOSFETs. Thermal annealing can recover the on/off characteristics of neutron-irradiated 4H-SiC MOSFETs. After thermal annealing, the interface-trap density decreased and the effective mobility increased in terms of the on-characteristics. This finding could be due to the improvement of the interfacial state from thermal annealing and the reduction in Coulomb scattering due to the reduction in interface traps. Additionally, in terms of the off-characteristics, the thermal annealing resulted in the recovery of the breakdown voltage and leakage current. After the thermal annealing, the number of positive trapped charges at the MOSFET interface was decreased.