• Journal of Radiation Protection and Research
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• v.36 no.4
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• pp.200-208
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• 2011

The neutron fluence measurement and evaluation technology is very important for material irradiation test. The most essential technology in this study is the neutron irradiation evaluation method using a fluence monitor. The fluence monitors were fabricated with metal wires of the purity ${\geq}$ 99.9%, whose dimensions were 0.1mm diameter, about 3 mm length, and around 150-200 ${\mu}g$ mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. After irradiation tests, radiation activities were measured with the high purity germanium (HPGe) detector. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position.

• Journal of Radiation Protection and Research
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• v.29 no.4
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• pp.263-268
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• 2004

Using the Monte Carlo simulation, a study on the lion-proportionality of the prototype phoswich detector with $2'{\times}2'$ CSI(Tl) and plastic scintillator, which was made by KAERI, has been carried. The defector response functions (DRFs) calculated by simulations were compared with the experimental measurement on the $^{137}Cs\;and\;^{60}Co$. To precisely simulate the DRF for the phoswich, the CSI(Tl) non-proportionality was calculated using the electron response and the simplified electron cascade sequence for treating the photoelectric absorption event. The resulting DRFs of $^{137}Cs\;and\;^{60}Co$ sources obtained by simulations were compared with experiments for verification. For $^{137}Cs$, gamma-ray responses simulated by MCNP5 are generally good agreement with the measured ones. But the DRF of $^{60}Co$ does not match well with the results of experiment in the energy region below second peak due to the coincidence effect of two gamma-rays (1.17 MeV and 1.33 MeV). Through the analysis of the non-proportionality of CsI(Tl) in the prototype phoswich, the improved DRFs considering non-proportionality were produced and the simulation results were verified using the experimental measurements. However, to more precisely reproduce the DRF for the phoswich, further studies in relation to the electron channeling effect and the Doppler broadening effect of a scintillator are still needed as well as considering that effect of the transfer contribution.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• The Journal of Korean Society for Radiation Therapy
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• v.25 no.2
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• pp.145-151
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• 2013

Purpose: In this study, we analyzed how the dose change by field size effects on atomic number of shielding materials while using 6 MeV election beam. Materials and Methods: The parallel plate chamber is mounted in $25{\times}25cm^2$ the phantom such that the entrance window of the detector is flush with the phantom surface. phantom was covered laterally with aluminum, copper and lead which thickness have 5% of allowable transmission and then the doses were measured in field size $6{\times}6$, $10{\times}10$ and $20{\times}20cm^2$ respectively. 100 cGy was irradiated using 6 MeV electron beam and SSD (Source Surface Distance) was 100 cm with $10{\times}10cm^2$ field size. To calculate the photon flux, electron flux and Energy deposition produced after pass materals respectively, MCNPX code was used. Results: The results according to the various shielding materials which have 5% of allowable transmission are as in the following. Thickness change rate with field size of $6{\times}6cm^2$ and $20{\times}20cm^2$ that compared to the field size of $10{\times}10cm^2$ found to be +0.06% and -0.06% with aluminum, +0.13% and -0.1% with copper, -1.53% and +1.92% with lead respectively. Compare to the field size $10{\times}10cm^2$, energy deposition for $6{\times}6cm^2$ and $20{\times}20cm^2$ had -4.3% and +4.85% respectively without shielding material. With aluminum it had -0.87% and +6.93% respectively and with lead it had -4.16% and +5.57% respectively. When it comes to photon flux with $6{\times}6cm^2$ and $20{\times}20cm^2$ of field sizes the chance -8.95% and +15.92% without shielding material respectively, with aluminum the number -15.56% and +16.06% respectively and with copper the chance -12.27% and +15.53% respectively, with lead the number +12.36% and -19.81% respectively. In case of electron flux in the same condition, the number -3.92% and +4.55% respectively without shielding material respectively, with aluminum the number +0.59% and +6.87% respectively, with copper the number -1.59% and +3.86% respectively, with lead the chance -5.15% and +4.00% respectively. Conclusion: In this study, we found that the required thickness of the shielding materials got thinner with low atomic number substance as the irradiation field is increasing. On the other hand, with high atomic number substance the required thickness had increased. In addition, bremsstrahlung radiation have an influence on low atomic number materials and high atomic number materials are effected by scattered electrons.