• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3824-3832
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• 2022

Prompt capture gammas are an important part of the fission reactor gamma field. Because some of the structural materials after neutron capture can emit photons with high energies forming the dominant component of the gamma spectrum in the high energy region, the following study of the high energy capture gamma was carried out. High energy gamma radiation may play a major role in areas of the radiation sciences as reactor dosimetry. The HPGe measurements and calculations of the high-energy aluminum capture gamma were performed at two moderator levels in the VR-1 pool-type reactor. The result comparison for nominal levels was within two sigma uncertainties for the major 7.724 MeV peak. A larger discrepancy of 60% was found for the 7.693 MeV peak. The spectra were also measured using a stilbene detector, and a good agreement between HPGe and stilbene was observed. This confirms the validity of stilbene measurements of gamma flux. Additionally, agreement of the wide peak measurement in 7-9.2 MeV by stilbene detector shows the possibility of using the organic scintillators as an independent power monitor. This fact is valid in these reactor types because power is proportional to the thermal neutron flux, which is also proportional to the production of capture gammas forming the wide peak.

• Analytical Science and Technology
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• v.5 no.1
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• pp.57-61
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• 1992

For the neutron spectrum measurement, $^6Li$ neutron spectrometer system is installed. The characteristic of the $^6Li$ detector are investigated using a $^{137}Cs$ and $^{207}Bi$ point source, and the neutron capture peaks and the pulse height spectrum using an $^{214}Am-Be$ neutron source are measured. Furthermore, the pulse height spectrum for the irradiation time variation from the (214)^Am-Be neutron source, and for the distance variation between detector and source, and the threshold variation of discriminator are measured.

• 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.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.626-636
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• 2021

Boron-neutron capture therapy (BNCT) is a cancer treatment method that exploits the high neutron reactivity of boron. Monitoring the prompt gamma rays (PGs) produced during neutron irradiation is essential for ensuring the accuracy and safety of BNCT. We investigate the imaging of PGs produced by the boron-neutron capture reaction through Monte Carlo simulations of a gamma camera with a SrI₂ scintillator and parallel-hole collimator. GAGG scintillator is also used for a comparison. The simulations allow the shapes of the energy spectra, which exhibit a peak at 478 keV, to be determined along with the PG images from a boron-water phantom. It is found that increasing the size of the water phantom results in a greater number of image counts and lower contrast. Additionally, a higher septal penetration ratio results in poorer image quality, and a SrI₂ scintillator results in higher image contrast. Thus, we can simulate the BNCT process and obtain an energy spectrum with a reasonable shape, as well as suitable PG images. Both GAGG and SrI₂ crystals are suitable for PG imaging during BNCT. However, for higher imaging quality, SrI₂ and a collimator with a lower septal penetration ratio should be utilized.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.165-175
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• 2015

The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.113-124
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• 2009

The line shape functions for the Doppler-broadened gamma ray spectrum are considered in the $^{10}B(n,{\alpha}{\gamma})^7Li$ reaction occurring in a surrounding medium where the excited $^7Li$ nucleus is slowed down and stopped before decay. The phenomenological form of the stopping power was used for the broadening effect. Convolution with the detailed response of a germanium detector is taken into consideration for the simplest case of solely electronic stopping. A numerical study for the analysis of $^{10}B$ by thermal neutron capture is conducted by performing a parametric search and fitting the measured spectrum in a least-squares approach. In comparison with the previous numerical approach using the same analysis, the computational speed is increased and reliable information concerning the stopping power of the medium is obtained while estimating the uncertainty. Implementation of the routine analysis of $^{10}B$ is facilitated on a recent version of the gamma ray spectrum analysis package HyperGam.

• Journal of radiological science and technology
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• v.43 no.5
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• pp.389-395
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• 2020

The evaluation of radioactivated components of heavy-ion accelerator facilities affects the safety of radiation management and the exposure dose for workers. and this is an important issue when predicting the disposal cost of waste during maintenance and dismantling of accelerator facilities. In this study, the FLUKA code was used to simulate the proton treatment device nozzle and classify the radio-nuclides and total radioactivity generated by each component over a short period of time. The source term was evaluated using NIST reference beam data, and the neutron flux generated for each component was calculated using the evaluated beam data. Radioactive isotopes caused by generated neutrons were compared and evaluated using nuclide information from the International Radiation Protection Association and the Korea Radioisotope association. Most of the nuclides produced form of beta rays and electron capture, and short-lived nuclides dominated. However, In the case of ⁵⁴Mn, which is a radioactive product of iron, the effect of gamma rays should be considered. In the case of tritium generated from a material with a low atomic number, it is considered that handling care should be taken due to its long half-life.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2717-2724
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• 2020

Silicon, especially silicon in the form of SiO₂, is a major component of rocks. Final spent fuel storages, which are being designed, are located in suitable rock formations in the Earth's crust. Reduction of the uncertainty of silicon neutron scattering and capture is needed; improved silicon evaluations have been recently produced by the ORNL/IAEA collaboration within the INDEN project. This paper deals with the nuclear data validation of that evaluation performed at the LR-0 reactor by means of critical experiments and measurement of reaction rates. Large amounts of silicon were used both as pure crystalline silicon and SiO₂ sand. The critical moderator level was measured for various core configurations. Reaction rates were determined in the largest core configuration. Simulations of the experimental setup were performed using the MCNP6.2 code. The obtained results show the improvement in silicon cross-sections in the INDEN evaluations compared to existing evaluations in major libraries. The new Thermal Scattering Law for SiO₂ published in ENDF/B-VIII.0 additionally reduces the discrepancy between calculation and experiments. However, an unphysical peak is visible in the neutron spectrum in SiO₂ obtained by calculation with the new Thermal Scattering Law.

• Progress in Medical Physics
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• v.25 no.3
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• pp.151-156
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• 2014

The purpose of this study was to confirm the feasibility of imaging of therapy region from the boron neutron capture therapy (BNCT) using the measurement of the prompt gamma ray depending on the neutron flux. Through the Monte Carlo simulation, we performed the verification of physical phenomena from the BNCT; (1) the effects of neutron according to the existence of boron uptake region (BUR), (2) the internal and external measurement of prompt gamma ray dose, (3) the energy spectrum by the prompt gamma ray. All simulation results were deducted using the Monte Carlo n-particle extended (MCNPX, Ver.2.6.0, Los Alamos National Laboratory, Los Alamos, NM, USA) simulation tool. The virtual water phantom, thermal neutron source, and BURs were simulated using the MCNPX. The energy of the thermal neutron source was defined as below 1 eV with 2,000,000 n/sec flux. The prompt gamma ray was measured with the direction of beam path in the water phantom. The detector material was defined as the lutetium-yttrium oxyorthosilicate (Lu0,6Y1,4Si0,5:Ce; LYSO) scintillator with lead shielding for the collimation. The BUR's height was 5 cm with the 28 frames (bin: 0.18 cm) for the dose calculation. The neutron flux was decreased dramatically at the shallow region of BUR. In addition, the dose of prompt gamma ray was confirmed at the 9 cm depth from water surface, which is the start point of the BUR. In the energy spectrum, the prompt gamma ray peak of the 478 keV was appeared clearly with full width at half maximum (FWHM) of the 41 keV (energy resolution: 8.5%). In conclusion, the therapy region can be monitored by the gamma camera and single photon emission computed tomography (SPECT) using the measurement of the prompt gamma ray during the BNCT.