• Title/Summary/Keyword: Cf-252

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Anisotropy and Dose Equivalents Conversion Factors for the Unmoderated $^{252}Cf$ Source (비감속 $^{252}Cf$ 중성자선원에 대한 비등방성교정인자 및 선량당량환산인자)

  • Jeong, Deok-Yeon;Chang, Si-Young;Yoon, Suk-Chul;Kim, Jong-Soo
    • Journal of Radiation Protection and Research
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    • v.18 no.2
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    • pp.71-79
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    • 1993
  • Form the pure Maxwellian distribution(kT= 1.42MeV), the effects upon calibration factors of encapsulating a $^{252}Cf$ spontaneous fission neutron source were investigated to establish a standard neutron field in the Secondary Standard Dosimetry Laboratory at Korea Atomic Energy Research Institute(KAERI). A Monte Carlo code MCNP was used in simulating the encapsulation SR-Cf-100 and SR-Cf-1273 to be real conditions. The anisotropy(FI) and fluence-to-dose equivalents conversion factors$(H/{\Phi})$ were evaluated and compared with other results. As the results, the FI was determined to be 1.061 at ${\theta}=90^{\circ}$ with ${\pm}0.2%$ statistical error and the $(H/{\Phi})$ was evaluated to be $333.9 [pSv\;cm^2]\;with\;{\pm}0.5%$ statistical error, which is lower by 1.8% than that recommended by the ISO 8529. This means physically that the neutron spectrum of the unmoderated $^{252}Cf$ source in KAERI is a little more softened than that by the ISO.

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Evaluation of Neutron Detection Efficiency of the Unified Non-Destructive Assay Using MCNPX Code (MCNPX 코드를 이용한 통합비파괴측정장치의 중성자 검출 효율 평가)

  • Won, Byung-Hee;Seo, Hee;Lee, Seung Kyu;Park, Se Hwan;Kim, Ho Dong
    • Journal of Radiation Protection and Research
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    • v.38 no.4
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    • pp.172-178
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    • 2013
  • In this study, neutron detection efficiency of the UNDA system, which has been developed for study on nuclear material accountancy in a future pyro-process facility, was evaluated by using the MCNPX code. The detection efficiency was evaluated as a function of (1) positions of $^{252}Cf$ neutron source in the axial and radial directions, and (2) thicknesses and locations of the container filled with the depleted uranium materials for two different designs of the UNDA. In the case of $^{252}Cf$ source positions, detection efficiency was distributed from 6.83% to 13.35%. As $^{252}Cf$ source was positioned at upper part in the axial direction, detection efficiency was decreased after a slight increase. On the other hands, as $^{252}Cf$ source was positioned at outer part in the radial direction, detection efficiency was increased. In the case of container thickness, there was a slight decline when the thickness was increased. As the container was located at upper part, detection efficiency was decreased and as the container was located at outer part, detection efficiency was increased. Detection efficiency was varied from 10.31% to 13.61%. These values were higher than that of $^{252}Cf$ source case. The UNDA with polyethylene cover has about 2% higher detection efficiency than the UNDA without the cover.

Development of Neutron Induced Prompt γ-ray Spectroscopy System Using 252Cf (252Cf 선원을 이용한 즉발감마선 계측시스템 구성)

  • Park, Yong-Joon;Song, Byung-Chul;Jee, Kwang-Yong
    • Analytical Science and Technology
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    • v.16 no.1
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    • pp.12-24
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    • 2003
  • For the design and set-up of neutron induced prompt ${\gamma}$-ray spectroscopy system using $^{252}Cf$ neutron source, the effects of shielding and moderator materials have been examined. The $^{252}Cf$ source being used for TLD badge calibration in Korea Atomic Energy Research Institute was utilized for this preliminary experiment. The ${\gamma}$-ray background and prompt ${\gamma}$-ray spectrum of the sample containing Cl were measured using HPGe (GMX 60% relative efficiency) located at the inside of the system connected to notebook PC at the outside of the system (about 20 meter distance). The background activities of neutron and ${\gamma}$-rays were measured with neutron survey meter as well as ${\gamma}$-ray survey meters, respectively and the system was designed to minimize the activities. Prompt ${\gamma}$-ray spectrum was measured using ${\gamma}$-${\gamma}$ coincident system for reduce the background and the continuum spectrum. The optimum system was designed and set up using the experimental data obtained.

Neutron Dose Measurements Using TLDs in a 252Cf Neutron Field (252Cf 중성자장에서 열형광선량계(TLD)를 이용한 중성자 방사선량 측정)

  • Chang, Insu;Kim, Sang In;Lee, Jung Il;Kim, Jang Lyurl;Kim, Bong Hwan
    • Journal of Radiation Protection and Research
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    • v.38 no.1
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    • pp.37-43
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    • 2013
  • In case of neutron dose measurement using TLDs (thermo-luminescence dosimeters), because the neutron energy dependence of the TLD is very high, the calibration of the energy response according to the characteristics of the neutron spectrum of workplace is required. In the present study, the ambient dose equivalent rates inside and around the Long-Counter (neutron detector) with narrow and complex inside in the neutron field of $^{252}Cf$ were evaluated. The calibration factors to account for the neutron energy dependence of TLDs were established for both the bare and $D_2O$ modulated $^{252}Cf$ neutron beams, respectively. The values of the TLD's measurement were compared with the computational results of the MCNPX (Monte Carlo N-Particles transport code). When using the two calibration factors of the TLD than a single calibration factor, the measured and the calculated values at the point of verification outside and inside the Long-Counter were in more good agreement. This results show that TLD should be calibrated in the reference neutron field similar to workplace situation.

Study on Thermal Neutron Efficiency for Neutron Induced Prompt Gamma-ray Spectrometer Using Various Reflectors (즉발감마선 계측시스템의 반사체를 이용한 열중성자 효율증대 연구)

  • Park, Y.J.;Song, B.C.;Jee, K.Y.
    • Analytical Science and Technology
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    • v.16 no.5
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    • pp.426-429
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    • 2003
  • Neutron induced prompt gamma-ray spectroscopy (NIPS) system equipped with a $^{252}Cf$ neutron source and a n-type coaxial HPGe detector was installed for the quantitative analysis of aqueous samples in KAERI, Korea. Since the thermal neutron flux for the $^{252}Cf$ neutron source is relatively low compared to that for the reactor, the use of a thermal neutron reflector in the NIPS system may lead to improved results. The enhancement by using various reflectors was carried out by comparing the Cl peak with or without a cadmium plate between sample and the $^{252}Cf$ source. The use of pyrolitic graphite as a reflector provided a good result.

Comparison Study of Experimental Neutron Room Scattering Corrections with Theoretical Corrections in RCL's Calibration Facility at KAERI (한국원자력연구소 중성자교정실에 대한 중성자산란보정인자 결정연구)

  • Yoon, Suk-Chul;Chang, Si-Young;Kim, Jong-Soo;Kim, Jang-Lyul;Kim, Bong-Hwan
    • Journal of Radiation Protection and Research
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    • v.22 no.1
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    • pp.29-33
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    • 1997
  • Neutron room scattering corrections that should be made when neutron detectors are calibrated with a $D_2O$ moderated $^{252}Cf$ neutron source in the center of a calibration room are considered. Such room scattering corrections are dependent on specific neutron source type, detector type, calibration distance, and calibration room configuration. Room scattering corrections for the responses of a thermoluminescence dosimeter and two different types of spherical detectors to neutron source in the Radiation Calibration Laboratory(RCL) neutron calibration facility at the Korea Atomic Energy Research Institute(KAERI) were experimentally determined and are presented. The measured room scattering results are then compared with theoretical results calculated by predicting room scattering effects in terms of parameters related to the specific configuration. Agreement between measured and calculated scattering correction is generally about 10% for three kinds of detectors in the calibration facility.

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The Development Study of A Manganese Sulphate Bath System ($MnSO_4$용액조 장치 개발 연구)

  • Hwang, Sun-Tae;Lee, Kyung-Ju;Choi, Kil-Oung;Kim, Won-Sik
    • Journal of Radiation Protection and Research
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    • v.11 no.1
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    • pp.70-76
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    • 1986
  • In order to establish the national standards of neutron measurements, a manganese sulphate ($MnSO_4$) bath system was developed under the IAEA technical support. This bath system was made up of a spherical s.s. 316 L bath, of 3.5 mm thick and of 125 cm internal diagmeter, filled with a manganese sulphate solution, a solution circulating system, and a $^5Mn\;{\gamma}-ray$ monitoring system. The solution pumped from the bath was introduced into a Marinelli beaker-type monitor vessel which was equipped with two seperate detectors, $3.8cm{\phi}{\times}3.8cm$ NaI(T1) crystals. The performance of the system were tested using the neutron sources, $^{241}Am-Be\;and\;^{252}Cf$, mounted at the center of the bath. From the decay curve analysis of $^{56}Mn$ activity, neutron emission rate of $^{252}Cf$ by the comparative method was obtained to be $3.71{\times}10^7\;n/s\;per\;50{mu}g$ as of November 15, 1985.

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Comparative optimization of Be/Zr(BH4)4 and Be/Be(BH4)2 as 252Cf source shielding assemblies: Effect on landmine detection by neutron backscattering technique

  • Elsheikh, Nassreldeen A.A.
    • Nuclear Engineering and Technology
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    • v.54 no.7
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    • pp.2614-2624
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    • 2022
  • Monte Carlo simulations were used to model a portable Neutron backscattering (NBT) sensor suitable for detecting plastic anti-personnel mines (APMs) buried in dry and moist soils. The model consists of a 100 MBq 252Cf source encapsulated in a neutron reflector/shield assembly and centered between two 3He detectors. Multi-parameter optimization was performed to investigate the efficiency of Be/Zr(BH4)4 and Be/Be(BH4)2 assemblies in terms of increasing the signal-to-background (S/B) ratio and reducing the total dose equivalent rate. The MCNP results showed that 2 cm Be/3 cm Zr(BH4)4 and 2 cm Be/3 cm Be(BH4)2 are the optimal configurations. However, due to portability requirements and abundance of Be, the 252Cf-2 cm Be/3 cm Be(BH4)2 NBT model was selected to scan the center of APM buried 3 cm deep in dry and moist soils. The selected NBT model has positively identified the APM with a S/B ratio of 886 for dry soils of 1 wt% hydrogen content and with S/B ratios of 615, 398, 86, and 12 for the moist soils containing 4, 6, 10, and 14 wt% hydrogen, respectively. The total dose equivalent rate reached 0.0031 mSv/h, suggesting a work load of 8 h/day for 806 days within the permissible annual dose limit of 20 mSv.

Estimation of Neutron Energy Spectrum of Cf-252 using Single Bonner Sphere with TLD-600 and TLD-700 (단일 보너구와 TLD-600 및 TLD-700을 이용한 Cf-252의 중성자 에너지 스펙트럼 평가)

  • Kim, Sunghwan;Cheon, Jongkyu;Lee, Jae Jin;Nam, Uk-Won
    • Journal of Sensor Science and Technology
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    • v.22 no.3
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    • pp.223-226
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    • 2013
  • We designed a single polyethylene bonner sphere with several thermo-luminescence dosimeters (TLD), for measurement of neutron energy spectrum. For the separation of the neutron dosage in the neutron-gamma mixed field, we used 21 ea TLD-600s and TLD-700s, respectively. Because, TLD-600 is sensitive to neutron and gamma rays, and, TLD-700 is sensitive only to gamma-rays, we could determine the each dose by neutron and gamma rays. The neutron response function of the bonner sphere with TLDs was calculated by MCNPX (ver. 2.5.0) Monte Carlo simulation in the energy range from $10^{-1}$ to 20 MeV. For the Cf-252 standard neutron source in KRISS, we could estimate the neutron energy spectrum by unfolding method using the response function.