• 대한방사선방어학회:학술대회논문집
• /
• 2011.11a
• /
• pp.312-313
• /
• 2011

• 대한방사선방어학회:학술대회논문집
• /
• 2011.04a
• /
• pp.36-37
• /
• 2011

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2011.05a
• /
• pp.403-404
• /
• 2011

• Progress in Medical Physics
• /
• v.18 no.2
• /
• pp.87-92
• /
• 2007

A thermal neutron beam facility utilizing a typical tangential beam port for Neutron Capture Therapy was installed at the HANARO, 30 MW multi-purpose research reactor. Mixed beams with different physical characteristics and relative biological effectiveness would be emitted from the BNCT irradiation facility, so a quantitative analysis of each component of the mixed beams should be performed to determine the accurate delivered dose. Thus, various techniques were applied including the use of activation foils, TLDs and ionization chambers. All the dose measurements were perform ed with the water phantom filled with distilled water. The results of the measurement were compared with MCNP4B calculation. The thermal neutron fluxes were $1.02E9n/cm^2{\cdot}s\;and\;6.07E8n/cm^2{\cdot}s$ at 10 and 20 mm depth respectively, and the fast neutron dose rate was insignificant as 0.11 Gy/hr at 10 mm depth in water The gamma-ray dose rate was 5.10 Gy/hr at 20 mm depth in water Good agreement within 5%, has been obtained between the measured dose and the calculated dose using MCNP for neutron and gamma component and discrepancy with 14% for fast neutron flux Considering the difficulty of neutron detection, the current study support the reliability of these results and confirmed the suitability of the thermal neutron beam as a dosimetric data for BNCT clinical trials.

• Journal of Radiation Protection and Research
• /
• v.12 no.2
• /
• pp.19-27
• /
• 1987

In order to determine the $^{56}Mn\;{\gamma}$-detection efficiency of a $MnSO_4$ bath system, it is essential to do the absolute activity measurement of $^{56}Mn$ solution. For the fabrication of $^{56}Mn$ samples, a 13.718 mg of $^{56}Mn$ metal flake with 99.99% purity was irradiated for 12 minutes at the thermal neutron field of about $10^{13}n/cm^2s$ of flux density. The neutron activated $^{56}Mn$ metal sample was dissolved in 50 ml of 0.1 N-HCl solution. The $^{56}Mn$ samples were fabricated by using the dissolved stock solution and the activity of each of them was measured by the $4{\pi}{\beta}-{\gamma}$ coincidence counting technique. The obtained result was 408.070 kBq/mg with total uncertainty of 0.366% at reference date, 0 h on October 15, 1987.

• Nuclear Engineering and Technology
• /
• v.23 no.3
• /
• pp.352-362
• /
• 1991

Recent popular trends in pressurized water reactor(PWR) fuel management are to extend the cycle length and to employ the low-leakage core designs for the optimal utilization of the uranium resources. In control strategy incorporated with the fuel management, turnable absorbers are required to control the power peaking and to ensure a negative moderator temperature coefficient during reactor operation. In this study, the nuclear characteristics and the optimal allocation of gadolinium-poisoned rods within the fuel assembly are considered using KWU SAV 79 A Code Package. First, analyses are carried out to compare the nuclear characteristics of the fuel assemblies contain-ing WABA(Wet Annular Burnable Absorber) and Gadolinium burnable absorbers respectively. The analyses show that the gadolinium-bearing fuel assembly has peculiar depletion characteristics ensuing from the very large thermal neutron absorption cross section. Peculiar characteristics of gadolinium provide basis for the optimal allocation of Gd rods in fuel assembly. Second, the methodology of an optimal allocation of gadolinium-poisoned rods within the fuel assembly is developed and applied to some nuclear designs.

• Journal of Sensor Science and Technology
• /
• v.18 no.5
• /
• pp.402-407
• /
• 2009

In this study, plastic optical fiber sensors which can measure thermal neutron dose in a mixed neutron-gamma field are developed and characterized. Using $^{252}Cf$ and $^{60}Co$ sources, the scintillators suitable for thermal neutron detection, are tested and the scintillating lights generated from a plastic optical fiber sensor in the Kyoto University Critical Assembly (kuca) core are measured. Also, the distributions of thermal neutron and gamma-ray are measured in a mixed field as a function of the distance from the center of the reactor core at KUCA and the distribution of thermal neutron is obtained using a subtraction method. Sensitivity of the fiber-optic radiation sensor system is about 0.49 V/mW according to power of the KUCA core and its relative error is about 1.2 %.