• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.93-99
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• 2001

Voxel head phantom for overcoming the limitation of mathematical phantom in depleting anatomical details was constructed and example dose calculation for BNCT was performed. The repeated structure algorithm of the general purpose Monte Carlo code, MCNP4B was applied for yokel Monte Carlo calculation. Simple binary yokel phantom and combinatorial geometry phantom composed of two materials were constructed for validating the voxel Monte Carlo calculation system. The tomographic images of VHP man provided by NLM(National Library of Medicine) were segmented and indexed to construct yokel head phantom. Comparison of doses for broad parallel gamma and neutron beams in AP and PA directions showed decrease of brain dose due to the attenuation of neutron in eye balls in case of yokel head phantom. The spherical tumor volume with diameter, 5cm was defined in the center of brain for BNCT dose calculation in which accurate 3 dimensional dose calculation is essential. As a result of BNCT dose calculation for downward neutron beam of 10keV and 40keV, the tumor dose is about doubled when boron concentration ratio between the tumor to the normal tissue is $30{\mu}g/g$ to $3{\mu}g/g$. This study established the voxel Monte Carlo calculation system and suggested the feasibility of precise dose calculation in therapeutic radiology.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1204-1212
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• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.79-86
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• 2015

$^3He$ gas has been used for neutron monitors as the neutron converter owing to its advantages such as high sensitivity, good ${\gamma}$-discrimination capability, and long-term stability. However, $^3He$ is becoming more difficult to obtain in last few years due to a global shortage of $^3He$ gas. Accordingly, the cost of a neutron monitor using $^3He$ gas as a neutron converter is becoming more expensive. Demand on a neutron monitor using an alternative neutron conversion material is widely increased. $^{10}B$ has many advantages among various $^3He$ alternative materials, as a neutron converter. In order to develop a neutron converter using $^{10}B$ (actually $B_4C$), we calculated the optimal thickness of a neutron converter with a Monte Carlo simulation using MCNP6. In addition, a neutron converter was fabricated by the Ar sputtering method and the neutron signal detection efficiencies were measured with respect to various thicknesses of fabricated a neutron converter. Also, we developed a 2-dimensional multi-wire proportional chamber (MWPC) for neutron beam profile monitoring using the fabricated a neutron converter, and performed experiments for neutron response of the neutron monitor at the 30 MW research reactor HANARO at the Korea Atomic Energy Research Institute. The 2-dimensional MWPC with boron ($B_4C$) neutron converter was proved to be useful for neutron beam monitoring, and can be applied to other types of neutron imaging.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1563-1570
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• 2017

In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP) code. The dose rate for the dual-purpose cask utilizing the recently developed materials of $epoxy/clay/B_4C$ and $epoxy/clay/B_4C/carbon$ fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of $epoxy/clay/B_4C$ instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.

• Journal of the Korean Society for Nondestructive Testing
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• v.14 no.4
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• pp.228-236
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• 1995

In this paper the method to evaluate shielding deficiency by gamma scanning test was presented and verified theoretically by Monte Carlo code which is one of the best effective method for radiation shielding calculation. The cylindrical shielding model was selected to evaluate shielding deficiency by gamma scanning test. First, the reference shielding according to the design requirement of cask was fabricated specially and reference values were measured with Co-60 source and scintillation detector. As a result with which calculated the reference values, it is shown that maximum deficiency thickness for lead of true cylindrical shielding model was 12mm. To verify this, thickness of lead was calculated by MCNP code and maximum deficiency thickness was 11.6mm. The experimental result obtained by the use of reference shielding was in good agreement with the theoretical result within 4.1%. So, this method can be applied to inspect the shielding ability for great shielding or cask which the radioactive material is used. To perform measurement more exactly, the further work on the development of measuring equipment to display the results on the screen will be required.

• 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.

• Progress in Medical Physics
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• v.18 no.2
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• pp.87-92
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• 2007

A thermal neutron beam facility utilizing a typical tangential beam port for Neutron Capture Therapy was installed at the HANARO, 30 MW multi-purpose research reactor. Mixed beams with different physical characteristics and relative biological effectiveness would be emitted from the BNCT irradiation facility, so a quantitative analysis of each component of the mixed beams should be performed to determine the accurate delivered dose. Thus, various techniques were applied including the use of activation foils, TLDs and ionization chambers. All the dose measurements were perform ed with the water phantom filled with distilled water. The results of the measurement were compared with MCNP4B calculation. The thermal neutron fluxes were $1.02E9n/cm^2{\cdot}s\;and\;6.07E8n/cm^2{\cdot}s$ at 10 and 20 mm depth respectively, and the fast neutron dose rate was insignificant as 0.11 Gy/hr at 10 mm depth in water The gamma-ray dose rate was 5.10 Gy/hr at 20 mm depth in water Good agreement within 5%, has been obtained between the measured dose and the calculated dose using MCNP for neutron and gamma component and discrepancy with 14% for fast neutron flux Considering the difficulty of neutron detection, the current study support the reliability of these results and confirmed the suitability of the thermal neutron beam as a dosimetric data for BNCT clinical trials.

• Journal of the Korean Society of Radiology
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• v.2 no.3
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• pp.33-35
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• 2008

In the neutron capture experiment and calculation, the neutron absorption and scattering are very important. Especially these effects are conspicuous in the resonance energy region and below the thermal energy region. In the present study, we obtained energy dependent neutron absorption ratios of natural rhodium in energy region from 0.003 to 100 eV by MCNP-4B Code. The coefficients for neutron absorption was calculated for several types of thickness. In the lower energy region, neutron absorption is larger than higher region, because of large capture cross section (1/v). Furthermore it seems very different neutron absorption in the large resonance energy region. These results are very useful to decide the thickness of sample and shielding materials.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.215-224
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• 2001

In order to comply with revised national regulationson radiological protection and to implement recent nuclear data and dose conversion factors, KOPEC developed a new point kernel gamma and beta ray shielding analysis computer program. This new code, named VisualShield, adopted mass attenuation coefficient and buildup factors from recent ANSI/ANS standards and flux-to-dose conversion factors from the International Commission on Radiological Protection (ICRP) Publication 74 for estimation of effective/equivalent dose recommended in ICRP 60. VisualShieid utilizes graphical user interfaces and 3-D visualization of the geometric configuration for preparing input data sets and analyzing results, which leads users to error free processing with visual effects. Code validation and data analysis were performed by comparing the results of various calculations to the data outputs of previous programs such as MCNP 4B, ISOSHLD-II, QAD-CGGP, etc.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.