• 전기의세계
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• v.19 no.4
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• pp.18-23
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• 1970

A neutron detector, in general, can not be utilized as the thermal neutron detecting chamber in the nuclear power reactor, especially P.W.R. due to the characteristics of high temperature, high pressure and high neutron flux in a reactor vessel. We have performed an experiment to detect the thermal neutrons at 400.deg. C and high flux of thermal neutron in a power reactor. Coating boron-10 on the aluminium plates by means of surface diffusion method at 600.deg. C for 5 hours in an electric furace, also we made a typical chamber which was compensated ionization chamber filled with free air as an ionization gas. It was checked the chamber characteristics in the TRIGA MARK-II Reactor at the power level from zero to 250KW. The chamber current showed a perfect linear increase to power increase. However, many variation of the measured current were observed within the power of 50KW.

• Journal of the Korean Chemical Society
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• v.7 no.1
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• pp.13-16
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• 1963

Thermal neutron activation analysis was applied to determine the trace amount of Vanadium and Manganese in Buyo and Kumsan Ginseng. These elements have been regarded to have great nutritional value and one of the indispensable factor in the growth of ginseng. The TRIGA MARK II Reactor in Atomic Energy Research Institute was used for the neutron source. The samples were irradiated for 10 minutes for Vanadium and for 5 minutes for Manganese at the neutron flux of about $1.28{\times}10^{12}n/cm^2/sec$ and the RCL 256 Channel Pulse-Height Analyzer connected with $2"{\times}2"$ Nal(Tl) was used for activity determination. The amounts were about 0.02 ppm for Vanadium and 20 ppm for Manganese, and it was also found that the amounts of the elements were slightly different depending on the kinds of ginsengs.

• 전기의세계
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• v.10
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• pp.29-37
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• 1963

일반적으로 이 함수의 측정은 원자로내에 Pile Oscillator봉이라고 칭하는 회전체를 넣어 이것에서 sinusoidal한 입력을 가할 수 있게 reactivity driving function을 얻는 것이 가장 중요한 역할을 하고 있는 관계로 이 봉의 회전속도변화(TRIGA MARK-II REACTOR 0.001-100cps, ARGONAUT REACTOR 0.01-20cps)를 매속 정확히 기록할 수 있어야 되는 것이다. 그러나 방사선으로 접근하기 어려운 원자로 근방에 있는 이 회전체의 속도측정은 원거리에서 취하여야 되는 고로 과거 사용하여온 110에서 25,000cpm을 측정할 수 있는 strobotac이나 100rpm에 대하여 2V을 발생시키는 tachometer등으로서는 어려운 문제가 되있는 것이다. 따라서 본연구에서는 이러한 점을 고려하여 binary switching circuit의 비선형적 특성의 해석과 이에 관계된 실험을 토시로한 pulse기술의 이용으로서 정밀도 0.0021에서 0.0043%를 갖는 정밀 회전속도측정기를 만들어 그 원리와 이용가치를 기술하고자 한다.

• 전기의세계
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• v.15 no.5
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• pp.1-5
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• 1966

The stability and safety of operation of a reactor is determined mainly by the sign and magnitude of its reactivity responses to temperature changes. Reactors are subject to temperature fluctuations due to the changes in reactor power and ambient temperature. These temperature fluctuations cause reactivity disturbances through changes in the nuclear and physical properties of the core. Because of these important phenomena by the temperature effects, a large portion of study and testing on a reactor design has been conducted. In this experiment the overall temperature coefficient of the TRIGA MARK-II reactor is measured. The basic procedure is to change the tgemperature of the water moderator, and from the movements of a newly recalibrated control rod(this is necessary due to the effects of fuel burn-up and control rod depression) required to mintain criticality, the reactivity worth of the temperature change is determined. From this measurement, the overall temperature coefficient seems to be smoothly varying, almost a linear function of temperature, and a value of approximately -0.267${\c}$/$^{\circ}C$ can be obtained for an average temperature range from $17.6^{\circ}C$ to $32.5^{\circ}C$.

• Journal of Radiation Protection and Research
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• v.7 no.1
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• pp.34-48
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• 1982

The prime purpose of this study is to realize an index quantity, absorbed dose index, defined by the ICRU for the characterization of ambient radiation level at any location for the purpose of radiation protection. The experiment has been designed to be carried out in two phases, namely, preliminary and main experiment. In the primary study a 30cm diameter sphere of polyethylene was used, while in the main experiment that of tissue equivalent material was fabricated and used. Both experiments were performed in the gamma-ray fields of $^{137}Cs\;and\;^{60}Co$, and in a neutron beam of thermal column of the TRIGA MARK-II research reactor. In the measurement of gamma-ray absorbed dose TLD-700 $(^{7}LiF)$ chips were used, and for the neutron dose both Au activation foils and TLD chips (TLD-600 $(^{6}LiF)$ and TLD-700 for the discrimination of gamma-ray contribution) were used. Theoretical assessment of the absorbed dose in the sphere phantom has been carried out in accordance with the Ehrlich's idea that deduced on the basis of Burlin's cavity theory in the case of gamma-ray irradiation. For the analysis of neutron dose fluence-KERMA rate conversion method was used. The explanation on the dose assessment is given in detail. Results obtained were numerically and statistically analyzed and the depth dose distributions are presented in the graphic forms with normalized values. In the concluding remarks, the possibility and difficulty of realizing the index quantity, including questions and problems to be solved are mentioned.

• Nuclear Engineering and Technology
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• v.1 no.2
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• pp.91-101
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• 1969

Resistivity changes of n-type float-zone silicon crystals with 6.4$\times$10$^{14}$ to 1.25$\times$10$^{17}$ phosphorus atoms/㎤ due to irradiation by (1) 1 MeV electrons, (2) two types of research reactors, and (3) $Co^{60}$ ${\gamma}$-ray sources were investigated. The results were analyzed on the basis of a simple exponential formula derived by Buehler. While the formula gave a fair fit in the low fluence range in most cases, the deviation was quite appreciable in the case of 1 MeV electron irradiation, and a linear change gave better fit in some cases. The large change in the carrier removal rate in electron-irradiated samples in the high fluence range was analyzed in detail in terms of the Fermi level cross-over of the defect levels. Based on the damage constants evaluated from the initial portion of data where the formula was applicable, the relative effectiveness of various radiation sources in causing the resistivity change in n-type silicon was compared. The TRIGA Mark II reactor neutrons, for example, were found to be about 40 times more effective than 1 MeV electrons. The dependence of the damage constant on the initial carrier concentration was also examined. The physical basis of the exponential law and the effect of the Fermi level cross-over of the defect levels on the resistivity change in the high fluence ranges are discussed.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.200-210
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• 2016

The neutron radiography reactor (NRAD) is a 250 kW Mark-II Training, Research, Isotopes, General Atomics (TRIGA) reactor at Idaho National Laboratory, Idaho Falls, ID, USA. The East Radiography Station (ERS) is one of two neutron beams at the NRAD used for neutron radiography, which sits beneath a large hot cell and is primarily used for neutron radiography of highly radioactive objects. Additional fuel elements were added to the NRAD core in 2013 to increase the excess reactivity of the reactor, and may have changed some characteristics of the neutron beamline. This report discusses characterization of the neutron beamline following the addition of fuel to the NRAD. This work includes determination of the facility category according to the American Society for Testing and Materials (ASTM) standards, and also uses an array of gold foils to determine the neutron beam flux and evaluate the neutron beam profile. The NRAD ERS neutron beam is a Category I neutron radiography facility, the highest possible quality level according to the ASTM. Gold foil activation experiments show that the average neutron flux with length-to-diameter ratio (L/D) = 125 is $5.96{\times}10^6n/cm^2/s$ with a $2{\sigma}$ standard error of $2.90{\times}10^5n/cm^2/s$. The neutron beam profile can be considered flat for qualitative neutron radiographic evaluation purposes. However, the neutron beam profile should be taken into account for quantitative evaluation.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4226-4230
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• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Journal of Radiation Protection and Research
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• v.24 no.2
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• pp.101-108
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• 1999

The influence of the relative humidity, the temperature and the velocity of supply air on evaporation rate has been studied with non-boiling forced evaporation system in order to treat very low level radioactive liquid wastes produced from the decontamination and decommissioning activities. Experimental data on the evaporation rate have been obtained with the divers variables and experimental equation of air velocity was also obtained by the correlation of those data. The decontamination factor of this system was also obtained by the experimental data from a simulated liquid waste containing Cs-137 radio isotope ; $DF=10^4$. Since the commercial system will be operated for the treatment of the very low level radioactive liquid waste produced from decontamination & decommissioning of TRIGA Mark-II&III research reactor, the environmental assessment has been conducted to improve the operational safety. Exposure dose rate for an individual member of general public was assessed, and it showed that it was very lower than individual dose limits. The release of radioactivity of radioisotope material (Cs-137) to the environment was assessed, and result showed that it was $4.637{\times}10^{-14}\;{\mu}Ci/cc$.

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