[en] So as to establish the biological effects of BNCT in the HANARO Reactor, biological damages in cells and animals with treatment of boron/neutron were investigated. And 124I-BPA animal PET image, analysis technology of the boron contents in the mouse tissues by ICP-AES was established. A Standard clinical protocol, a toxicity evaluation report and an efficacy investigation report of BNCT has been developed. Based on these data, the primary permission of clinical application was acquired through IRB of our hospital. Three cases of pre-clinical experiment for boron distribution and two cases of medium-sized animal simulation experiment using cat with verifying for 2 months after BNCT was performed and so the clinical demonstration with a patient was prepared. Also neutron flux, fast neutron flux and gamma ray dose of BNCT facility were calculated and these data will be utilized good informations for clinical trials and further BNCT research. For the new synthesis of a boron compound, o-carboranyl ethylamine, o-carboranylenepiperidine, o-carboranyl-THIQ and o-carboranyl-s-triazine derivatives were synthesized. Among them, boron uptake in the cancer cell of the triazine derivative was about 25 times than that of BPA and so these three synthesized methods of new boron compounds were patented

[en] 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 1.19x10⁹ n/cm²s and 152, respectively, at 24 MW reactor power. The measured in-phantom neutron flux has a maximum value at a depth of 3mm 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

[en] The operational characteristics of the duoplasmatron ion source are investigated in order to obtain the maximum achievable extraction current of the He⁺ ion beam with the small divergence. Under the variations of the gas pressure, the arc current, the magnet current and the extraction voltage of the ion source, the change of the extracted He⁺ ion beam current is observed. An oxide filament, the mixture of BaO and SrO coated on Ni meshes, is used as the hot cathode, and its average lifetime is about 100 hours, The extraction current is linearly proportional to the arc current. An the magnet current of the ion source is increased, the extraction current increases, but the beam divergence becomes larger. The maximum extraction current is obtained at the source pressure of 0.084 Torr. The extraction current is proportional to the extraction voltage raised to the power of 3/2 as estimated from theory. At the extraction voltage of 5.72 kV, the maximum extraction current of 50 μA is obtained under the optimized extraction condition. (Author)

[en] Cs-137 with relatively long half-life is representative radionuclide fission product. Without nuclear fission, there are so few ways to produce the Cs-137 isotope. So, the detection of Cs-137 is very important in the evaluations of nuclear facility accident and waste disposal. Since Cs-137 emits the gamma-rays with almost unique energy of 661.660 keV, it is also very important isotope in the calibration of the gamma-ray detection system. Co-60 is also widely used isotope in the calibration of gamma-ray detection system, and it is produced in the significant amounts by neutron activation in nuclear reactor. It emits the prevailing two gamma-rays with energies of 1173.238 and 1332.502 keV which are larger than 1022 keV. Therefore, the single escape(SE) and double escape(DE) peaks can appear on the gamma-ray spectrum of Co-60. There are very few measurements of these escape peaks because of very low pair production probability of Co-60 gamma-rays. However, accidentally, the centroid energy of SE peak of 1173.238 keV gamma-rays is very near to that of gamma-ray peak from Cs-137. Considering the energy resolution of detection system, the uncertainty of energy calibration and the Doppler broadening of SE peak, it is highly probable that the SE peak of 1173 keV gammarays from Co-60 gives the misjudgement for the existence of the Cs-137 and the interference for peak area of Cs- 137 gamma-rays. In this work, we measured the Co-60 gamma-ray spectra with various full energy peak areas by using the HPGe detector, and we set the critical limit by which we can decide whether the net count of 662 keV peak would be significant or not

[en] The neutron flux is measured in the HANARO NTD2 irradiation hole after installing the irradiation facility. The Au wires, which are neutron flux monitors, are irradiated in the NTD2 irradiation hole. The radioactivity of irradiated Au wire is measured by gamma scanning method, and the neutron flux distribution is determined. The measured neutron flux distribution is very similar to the result of MCNP calculation used in the design of irradiation facility, so, the reliability of design calculation is verified. In the measurements with the fabricated irradiation tube without any modification, the neutron flux near the center of irradiated area is lower than that of edge, and the flux of upper region is higher than that of lower region. The deviation between the measurements is within 7%. So, the Al wire is wound round the neutron screen, and the flux is measured again. The neutron irradiation can be flattened within ±2.5%. Also, the variation of neutron irradiation according to the thickness of water gap inside of irradiation tube is investigated, and the maximum difference of 5.5% is occurred from the thickness of water gap of 3 mm. Conclusively, it is confirmed that the NTD service can be performed using the present irradiation facility

[en] By using this doping method, silicon semiconductors with an extremely uniform dopant distribution can be produced. They are usually used for high power devices such as thyristor (SCR), IGBT, IGCT and GTO. Now, the demand for high power semiconductor devices has increased rapidly due to the rapid increase of the green energy technologies. Among them, the productions of hybrid cars or fuel cell engines are excessively increased to reduce the amount of discharged air pollution substances, such as carbon dioxide which causes global warming. It is known that the neutron-transmutation-doped floating-zone (FZ) silicon wafers are used in insulated-gate bipolar transistors (IGBTs) which control the speed of the electric traction motors equipped in hybrid or fuel cell vehicles. Therefore, inevitably, it can be supposed that the demand of the NTD silicon is considerably increased. However, it is considered likely that the irradiation capacity will not be large enough to meet the increasing demand. After all, the large irradiation capacity for NTD such as a reactor dedicated to the silicon irradiation will be constructed depending on the industrial demand for NTD silicon. In this work, we investigated the relationship between the hybrid electric vehicle (HEV) industry and the NTD silicon production. Also, we surveyed the prospect for the production of the HEV. Then, we deduced the worldwide demand for the NTD silicon associated with the HEV production. This work can be utilized as the basic material for the construction of the new irradiation facility such as NTD-dedicated neutron source

No abstract available

[en] From the various analyses, it is confirmed that the failed fuel detection system(FFDS) of HANARO has some difficulties in monitoring the fuel failure because of the N-16. So, the FFDS based on the delayed neutron measurement is suggested, and the possibility for the failed fuel detection by using the neutron measurements is investigated in the present work. In normal operation of the reactor, the neutrons generated at the primary cooling circuit are assumed as the photoneutrons emitted by N-16 gamma rays and delayed neutrons by the fuel surface contamination, and the changes of neutron generations before and after reactor shutdown are calculated. From the comparison of calculations and measurements using BF3 detector, it is confirmed that about 70 % of the measured neutrons are delayed neutrons, and 30 % are photoneutrons. The neutron generation rate is calculated by Monte Carlo method, ard the calculated count rate of the photoneutron is 11 % of the whole measured neutron count rate. The amount of uranium exposed to the coolant in the core is also predicted by the delayed neutron transport calculation, and it is lower than the allowable level of surface contamination of the fuel rods. Conclusively, since the level of delayed neutrons generated from the surface contamination of fuel rods are sensitively monitored by the detection method of neutron in primary cooling circuit, it is confirmed that the detection method of delayed neutron is very useful for the monitoring of fuel failure

[en] While neutron absorption cross sections for a single energy (for example cross sections at 2,200m/s neutron velocity) can be quite unambiguous, much confusion has occurred in stating the effective cross section for a certain neutron spectrum. In considering the targets exposed to the neutron fields of a thermal reactor, thermal neutrons contribute an overwhelming majority of total reaction rate. Consequently the reaction rate or thermal flux measured using the usual 2,200m/s cross sections can be somewhat close to the actual values. As the reaction rate by epithermal neutrons will be different for every target material, the neutron fluxes measured under the same condition give different results, and the neutron fluxes measured with the same target also give different results according to the spectra. Westcott epithermal index, i.e,. the relative strength of the epithermal dE/E component in the neutron density including the thermal and epithermal neutrons, at the ENF beam port of HANARO are measured by the cadmium ratio method and the multi foil activation method. And effective absorption cross sections for Au-198(n,γ) and Co-60(n, γ) reactions and Westcott flux are determined

[en] During the normal operation of HANARO, 30 MW open-pool type research reactor, several radionuclides are dispersed in pool or coolant water. The nuclides in the water are generated by activations of coolant, dissolved substances and corrosion products. In addition, there are several fission products caused by uranium contamination on fuel cladding surface. The radionuclides disappear by radioactive decay, release out of the water, and they are also eliminated by the ion exchange resin in the coolant purification system. Thus, the information on its efficiency to remove the nuclides from the coolant is necessary to estimate the radiological effect of a research reactor. In this work, the efficiency is measured by the gamma-ray spectroscopy for the coolant of HANARO