• Geophysics and Geophysical Exploration
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• v.18 no.3
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• pp.154-161
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• 2015

For efficiently designing neutron induced gamma spectroscopy sonde, Monte Carlo simulation is employed to understand a dominant location of thermal neutron and classify the formation elements from the energy peak of capture gamma spectrum. A pulsed neutron generator emitting 14 MeV neutron particles was used as a source, and flux of thermal neutron was calculated from the twelve detectors arranged at each 10 cm intervals from the source. Design for reducing borehole effects using shielding materials was also applied to numerical sonde model. Moreover, principal elements and quantities of numerical earth models were verified through the energy spectrum analysis of capture gamma detected from a gamma detector. These results can help to enhance the signal-to-noise ratio, and determine an optimal placement of capture gamma detectors of neutron induced gamma spectroscopy sonde.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.115-120
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• 1997

영광3호기 방사선관리구역에 대한 중성자선량률을 정확히 평가하기 위하여 MCNP4A 전산코드를 이용, 방사선관리구역에서의 중성자 스펙트럼 계산을 수행하였다. 영광3호기에 대한 보다 정확하고 정밀한 3차원 몬테칼로 모델을 구축하기 위하여 핵연료집합체 구성요소 및 원자로심을 둘러싸고 있는 baffle, barrel,압력용기 등을 정확하게 묘사하였으며, 특히 방사선관리구역 주위의 구조물에 대해서도 3자원 MCNP 모델을 구축함으로써 원자로심부터 방사선관리구역까지 완전한 몬테칼로 모사(full-scope Monte Carlo simulation)를 이용한 계산을 수행하였다. 계산결과는 에너지 구간에 따른 중성자속 스펙트럼으로 나타내었으며 이 결과를 바탕으로 중성자속에 대한 선량률 환산인자를 고려하여 중성자선량률을 계산할 수 있다.

• Journal of the Korean Society of Radiology
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• v.1 no.1
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• pp.45-49
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• 2007

keV-neutron capture gamma-ray spectrum of $^{197}Au$(natural gold) sample have been measured in neutron energy range from 10 to 90 keV using the 3-MV pelletron accelerator of the Research Laboratory for Nuclear Reactors at the Tokyo Institute of Technology. Pulsed keV neutrons were produced from the $^7Li(p,n)^7Be$ reaction by bombarding on the $^7Li$ target with the 1.5-ns bunched proton beam. The incident neutron spectrum on the Au sample was measured by a $^6Li$-glass scintillation detector and TOF method. Capture gamma-rays from Au sample were measured by anti-Compton NaI(TI) spectrometer. Five average neutron energy regions were selected to obtain the neutron capture spectrum. Several gamma-ray peaks in the spectrum were found in the present experiment.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.373-373
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• 2002

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.11a
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• pp.248.1-248.1
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.447-447
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• 2003

• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.158-167
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• 2007

It is required that the neutron dosimetry be present to monitor the reactor vessel throughout its plant life. The Ex-vessel Neutron Dosimetry Systems which consist of sensor sets, radiometric monitors, gradient chains, and support hardware have been installed for 3-Loop plants after a complete withdrawal of all six in-vessel surveillance capsules. The systems have been installed in the reactor cavity annulus in order to characterize the neutron energy spectrum over the beltline region of the reactor vessel. The installed dosimetry were withdrawn and evaluated after a irradiation during one cycle and then compared to the cycle specific neutron transport calculations. The reaction rates from the measurement and calculation were compared and the results show good agreements each other.

• Nuclear Engineering and Technology
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• v.5 no.4
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• pp.334-338
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• 1973

The horizontal distribution of the fast neutron flux density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II reactor at the steady power of 250 KW has been measured using a solid state track detector which is natural mica placed in contact with $^{232}$ Th fissile foil. The neutron flux density was calculated on the assumption that the fast neutron spectrum is similar to that from the thermal-induced $^{235}$ U fission. The resulting flux density distribution along the horizontal line from the center of the thermalizing column door is presented in tabular and graphical forms.

• Journal of Radiation Protection and Research
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• v.7 no.1
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• pp.56-60
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• 1982

A calculation of the effective multiplication factor has been made for GODIVA and JEZEBEL critical assemblies by using a computer code, ANISN, with having the Watt's, Cranberg's and Maxwellian formulae for the prompt fission neutron spectrum as a fission source. Then the calculated values have been compared with experimental data obtained by others. The Maxwellian formula seems to be the best one for representing the prompt fission neutron spectrum since the effective multiplication factor based on it shows a better agreement with the experimental value compared to the rest formulae.

• Journal of Radiation Protection and Research
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• v.2 no.1
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• pp.17-23
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• 1977

The neutron dosimetrical parameters, i. e., the fission neutron spectrum-averaged cross-sections and the fluence-to-dose conversion factors have been calculated for some threshold detectors with the aid of a computer. The threshold detectors under investigation were the $^{115}In(n,\;n')^{115m}In,\;^{32}S(n,\;p)^{32}P$ and $^{27}Al(n,\;{\alpha})^{24}Na$ reactions. It is revealed that the average cross-sections($\bar{\sigma}$) for the $^{32}S(n,\;p)^{32}P$ reaction are independent of the spectral functions, namely, the Watt-Cranberg and Maxwellian forms. In the case of the $^{27}Al(n,\;{\alpha})^{24}Na$ reaction a variation of the $\bar{\sigma}$ values appears to be highly dependent on the fissioning types. It seems that both the average cross-section for the $^{115}In(n,\;n')^{11m}In$ reaction and the conversion factor are insensitive to the spectral deformation of fission neutrons. These phenomena make it applicable to use indium as a possible integral fast neutron dosimeter in nuclear criticality accidents provided that the virgin fission neutrons are completely free from the scattered neutrons.