• Journal of Radiation Protection and Research
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• v.10 no.1
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• pp.67-73
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• 1985

The manganous sulfate bath method for neutron source calibrations at the K-SRI is described together with the measurement of neutron emission rate of a source and the corrections applied for capture by competing nuclei of neutrons, and thermal neutron leakage, neutron absorption in the source itself. The commercially available neutron sources (Am-Be, $^{252}Cf$) for the calibration checks of neutron radiation instruments in the MeV range are considered in this paper.

• Proceedings of the KAIS Fall Conference
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• 2001.05a
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• pp.155-161
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• 2001

반도체 소자의 기판 재료로 사용되고 있는 실리콘 웨이퍼는 그 정밀도가 매우 중요하다. 본 연구에서는 균일한 Dopant 농도 분포를 얻을 수 있는 중성자 변환 Doping을 이용하여 실리콘에 인(P)을 Doping하는 연구를 수행하였다. 중성자 변환 Doping, 즉 NTD(Neutron Transmutation Doping)란 원자번호 30인 실리론 동위원소에 중성자가 조사되면 원자번호 31인 실리콘으로 변환되고, 2.6시간의 반감기를 갖고 decay 되면서 인(P)으로 변하게 되어 실리콘 웨이퍼에 n-type 전도를 갖게 하는 것을 말한다. 본 연구에서는 하나로 원자로를 이용하여 고저항(1000-2000Ω㎝) FZ 실리콘 웨이퍼에 중성자 조사하여 저항의 변화를 관찰하였고, 중성자 조사시 발생하는 점결함을 분석하여 점결함이 저항 변화에 미치는 영향을 알아보았다. 중성자 조사 전 이론적 계산에 의해 16.8Ω㎝와 4.76Ω㎝의 저항을 얻을 수 있을 것으로 예상되었고, 중성자 조사 후 SRP로 측정한 결과 실리콘 웨이퍼가 3Ω㎝과 2.5Ω㎝의 저항을 가지고 있을 확인할 수 있었으며, FT-IR 분석결과 점결함의 변화 양상을 확인할 수 있었다.

• Proceedings of the Korea Air Pollution Research Association Conference
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• 2003.11a
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• pp.229-230
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• 2003

중성자 방사화분석법은 핵반응을 통해 생성시킨 방사성 동위원소로부터 방출되는 방사선을 검색하여 절대적으로 성분원소를 정량하는 핵분석기술(Nuclear Analytical Techniques)이다. 현재 한국원자력연구소의 중성자 방사화분석실에서는 대기환경분야 응용연구로서 수년 동안 대기분진을 채집하여 미량 성분원소를 정량하고 있으며, 방법의 유효화와 측정결과의 신뢰성을 확보하기 위하여 시료의 채집 및 준비, 원소분석, 측정결과의 검증 등 분석과정에 대한 품질관리를 수행하고 있다. (중략)

• 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.

• Journal of Radiation Protection and Research
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• v.30 no.1
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• pp.35-37
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• 2005

The radiation detector research group at KAERI has developed a high efficiency neutron detector using a Gas Electron Multiplier (GEM). The double GEM was fabricated and operated in an Ar/Isobutane mixture. For an application to a high efficiency neutron detector, $^6Li\;or\;^{10}B$ neutron converters coated on each surface of the multi GEM foils were considered. The optimized thickness of the thin film for a neutron detection was calculated with the MCNP and SRIM. The neutron efficiency was calculated by changing the chemical components of the thin film, and the thickness of the thin film. The thermalized neutrons were measured by a GEM detector with a thin neutron converter on the drift plate.

• Nuclear Engineering and Technology
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• v.2 no.2
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• pp.85-95
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• 1970

An attempt has been made to do interpretation of the fast neutron dose with two threshold detectors incorporated with the Harwell criticality locket. This method is based on the assumption that the spectral distribution of fission neutrons in criticality accidents may be governed by one spectral parameter. The surface-absorbed dose for a unit fission neutron fluence seems to be insensitive to spectral shifts of the fission neutron spectrum. The average cross-sections for the activation detectors, however, are considerably changed with the neutron spectral shape, which may lead to a large error in calculating the dose from the reaction rate if one uses a fixed value for the average cross sections regardless of the neutron spectral distribution. Besides, the doses calculated from three representative formulae for fission neutron spectra have been compared : these formulae are Watt, Cranberg at al. and Maxwellian forms. The results obtained front the Maxwellian formula show a departure from the Watt and Cranberg's, both being similarly close.

• 대한방사선방어학회:학술대회논문집
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• 2011.04a
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• pp.118-119
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• 2011

• Journal of the Korean Society of Radiology
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• v.8 no.6
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• pp.287-292
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• 2014

In this study, the neutron capture cross section of $^{180}Ta$(natural existence ratio: 0.012 %) obtain by measuring has been compared with the evaluated data for the capture data. In generally, the neutron capture resonance is defined as Breit-Wigner formula. The formula consists of the resonance parameters such as neutron width, total width and neutron width. However in the case of $^{180}Ta$, these are very poor experimental neutron capture cross section data and resonance information in below 10 eV. Therefore, in the study, we analyzed the neutron resonance of $^{180}Ta$ with the measuring the prompt gamma-ray from the sample. And the resonance was compared with the evaluated data by Mughabghab, ENDF/B-VII, JEFF-3.1 and TENDL 2012. Neutron sources from photonuclear reaction with 46-MeV electron linear accelerator at Research Reactor Institute, Kyoto University used for cross section measurement of $^{180}Ta(n,{\gamma})^{181}Ta$ reaction. $BGO(Bi_4Ge_3O_{12})$ scintillation detectors used for measurement of the prompt gamma ray from the $^{180}Ta(n,{\gamma})^{181}Ta$ reaction. The BGO spectrometer was composed geometrically as total energy absorption detector.

• Nuclear Engineering and Technology
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• v.7 no.3
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• pp.191-198
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• 1975

A study was made on the neutron dosimetry in a mixed gamma-neutron field with LiF thermoluminescent dosimeter. In order to estimate the neutron dose in a mixed field, $^{6}$ LiF and $^{7}$ LiF dosimeters were used for fast and thermal neutron doses. The over-all conversion factors for the effects of dosimeter positions were derived for personnel monitoring and the glow curves of the LiF dosimeters for neutron and gamma-ray doses were also analyzed.

• 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.