• Journal of the Korean Society of Radiology
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• v.10 no.5
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• pp.337-341
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• 2016

We measured the neutron capture cross-section of natural Sm(n,${\gamma}$) reaction in the energy regions from 0.003 to 10 eV. The 46-MeV electron linear accelerator of Research Reactor Institute, Kyoto University was used for generating a continuous neutron source. The neutron time-of-flight method was adopted for energy measurement. An assembly of BGO($Bi_4Ge_3O_{12}$) scintillators composed of 12 pieces of BGO crystals measured prompt gamma rays from Sm(n,${\gamma}$) reaction. The BGO assembly was located at a distance of $12.7{\pm}0.02m$ from the neutron source. In order to determine the neutron flux impinging on the Sm, the $^{10}B(n,{\alpha}{\gamma})^7Li$ standard cross-section were used. Natural Sm(n,${\gamma}$) reaction measurement result of the neutron capture cross-section was compared with the results of evaluation of the BROND-2.2 and the previous experimental data of J. C. Chou and V. N. Kononov.

• The Journal of the Korea Contents Association
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• v.5 no.5
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• pp.133-139
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• 2005

The neutron capture cross section of $^{99}Tc$ has been measured relative to the $^{10}B(n,\gamma)$ standard cross section by the neutron time-of-flight(TOF) method in the energy range of 0.007 eV to 47 keV using a 46-MeV electron linear accelerator(linac) at the Research Reactor. Institute, Kyoto University(KURRI). In order to experimentally prove the result obtained, the supplementary cross section measurement has been made from 0.3 eV to 1 keV using the Kyoto University Lead stowing-down spectrometer (KULS) coupling to the linac. The relative measurement by the TOF method has been normalized to the reference value(20.01 b) at 0.0253 eV and the KULS measurement to that by the TOF method. The existing experimental data and the evaluated capture cross sections in ENDF/B-VI, JENDL-3.2, and JEF-2.2 have been compared with the current measurements by the linac TOF and the KULS experiments.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.12-17
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• 1989

The correction factor of neutron fraction absorbed by $^{55}$ Mn in the MnSO$_4$ bath was determined for the absolute measurement of neutron emission rate by using the solution circulation-type manganese sulfate bath system. For the determination of this correction factor, I/f, the atomic number desnsity and the effective neutron capture cross section data of Mn, S and impurity elements in the MnSO$_4$ solution were determined. For the atomic number density determination, the MnSO$_4$ solution concentration was determined by using the volumetric EDTA titration and gravimetric method. The impurity contents were analyzed by using the ICP method. For the calculation of effective neutron capture cross sections, a FORTRAN computer program EASCAL was developed in this study. in which Westcott's parameters and Axton's empirical relations are used.

• Journal of the Korean Society of Radiology
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• v.4 no.4
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• pp.29-32
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• 2010

The neutron capture cross section of $^{197}Au$ has been measured relative to the $^{10}B(n,{\gamma})$ standard cross section by the neutron time-of-flight(TOF) method using a 46-MeV electron linear accelerator(linac) at the Research Reactor Institute, Kyoto University(KURRI). In order to experimentally prove the result obtained, the supplementary cross section measurement has been made from 0.1 eV to 10 keV using the Kyoto University Lead slowing-down spectrometer (KULS) coupling to the linac. The relative measurement by the TOF method has been normalized to the reference value(24.5 b) at 1 eV. The evaluated capture cross sections in JENDL/D-99 Dosimetry have been compared with the current measurements by the KULS experiments.

• 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.

• The Journal of the Korea Contents Association
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• v.4 no.2
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• pp.76-82
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• 2004

The thermal neutron capture cross-section (at 2,200 m/s value) of the $^{141}$Pr(n,$\gamma$)$^{142}$Pr reaction was measured by an activation method by using the heavy water ($D_2$O) thermal neutron facility at the KUR(Kyoto University Reactor). The thermal neutron fiux used in this experiment was monitored with the$^{197}$Au(n,$\gamma$)$^{198}$Au standard cross-section. The previous results and the evaluated data of JENDL-3.2, ENDF/B-VI, and JEF-2.2 were in good agreement with the current result.

• Journal of the Korean Society of Radiology
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• v.17 no.1
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• pp.17-23
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• 2023

In this study, the thermoluminescence kinetics of electron trap in Li₆Y_0.5Gd_0.5(BO₃)₃ (LY_0.5G_0.5BO) scintillator for neutron detection composed of Li, Gd, and B with a high neutron response cross-section were investigated. The thermoluminescence glow curve of the LY_0.5G_0.5BO scintillation single crystal was measured and analyzed using the peak shape method, the initial rise method, and the machine learning algorithm to evaluate the physical parameters of the electron trap. The glow curve of the LY_0.5G_0.5BO scintillation single crystal consisted of a single peak. As a result of analyzing this peak, the activation energy, emission order, and frequency factor of the electron trap were 0.61 eV, 1.1, and 1.7×10⁷ s^-1, respectively. In addition, the possibility of thermoluminescence analysis of scintillators using machine learning was confirmed.

• Journal of the Korea Concrete Institute
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• v.28 no.2
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• pp.197-204
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• 2016

Concrete walls of neutron generating facilities such as fusion reactors and fission reactors become radioactive by neutron irradiation. Both low-activation and neutron shielding are a critical concern at the dismantling stage after the shutdown of facilities with a requirement of radioactive waste management. To tackle this, two types of additives were investigated in fabricating mortar specimens: synthetic high polymers and boron carbide. It is well known that a hydrogen atom is effective in neutron shielding by an elastic scattering because its mass is almost the same as that of the neutron. And boron is an effective neutron absorber with a big neutron absorption cross section. In this study, the effect of the type, shape, and size of polymers were investigated as well as that of boron carbide. Total 16 mix designs were prepared to reveal the effect of polymers on mechanical properties and neutron shielding performance. The neutron does equivalent of polymers-based mortar for fast neutrons decreased by 36 %, and the count rate of boron carbide-based mortar with regard to thermal neutrons decreased by 90 % compared to conventional mortar. These results showed that a combination of polymers and boron carbide compounds has potential to reduce the thickness of neutron shields as well as radioactive waste from reactors.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.119-125
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• 1997

PHWR achieves high neutron economy by adopting heavy water as its moderator and coolant. On the other hand it permits much tritium generation, compared to LWR, due to the neutron capture reaction of deuterium in heavy water. Meanwhile in the reactor core, $^3He formed as the result of-decay of tritium, captures a thermal neutron and transforms to tritium again. The existing calculation models on tritium generation in PHWR neglect the contribution of $^3He$ in both moderator and coolant due to its relatively low solubility. However the neutron capture cross-section of $^3He$ is almost $1.6{\times}10^7$ times as large as that of deuterium. That means that the dissolved amount of 0.03 ppm of $^3He$ in heavy water is enough to generate the same amount of tritium as that generated by the deuterium of total heavy water in the system. This study dealt with the contribution of $^3He$ to tritium generation. As a sample case, the contribution of $^3He$ to the tritium generation in Wolsong #1 was evaluated and compared to the measured values. According to the result of this study, it is concluded that $^3He$ in coolant contributes very much to the tritium generation but that in moderator shows negligible effects due to the low solubility and $^4He$ cover gas. At the beginning of the plant operation, the contribution of $^3He$ is slightly greater than the measured value but agrees well with the measured as the operating time increases.

• Nuclear Engineering and Technology
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• v.4 no.4
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• pp.340-344
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• 1972

Isomeric ratios were measured for the capture of thermal neutron by $^{79}$ Br, $^{80}$ Se, $^{103}$ Rh, $^{115}$ In and $^{133}$ Cs as well as those of epi-cadmium neutron by $^{79}$ Br, $^{80}$ Se and $^{l33}$Cs. The measurements were performed by analysing decay curves obtained by ${\gamma}$-ray spectrometry after irradiation. The counting efficiency curve was determined by using the calibrated standard sources with overall uncertainties of about 1%. Isomeric ratios, given in $\sigma$ high spin/($\sigma$ high spin + $\sigma$ low spin), of $^{80, 80m}$Br, $^{81,81m}$Se, $^{014, 104m}$Rh, $^{116,116m}$In and $^{134, 134m}$Cs produced by thermal neutron activation were found to be 0.21$\pm$0.01, 0.14$\pm$0.02, 0.12$\pm$0.02, 0.69$\pm$0.07 and 0.058$\pm$0.004, respectively, Those values of $^{80, 80m}$Br, $^{81,81m}$Se, and $^{134, 134m}$Cs Produced by epi-cadmium neutron were found to be 0.19$\pm$0.02, 0.29$\pm$0.02 and 0.074$\pm$0.011, respectively. The experimental values obtained were compared with the theoretical values deduced from the statistical model. There were the general agreements between the theory and the experiment.t.