• Advances in materials Research
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• v.4 no.1
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• pp.23-30
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• 2015

Energy absorption and exposure buildup factor have been computed for natural uranium in the energy range of 0.05-15MeV up to penetration depth of 40 mfp. Five-parameter geometric progression (G-P) fitting method has been used to compute buildup factors of uranium. The variation of energy absorption and exposure buildup factors with, penetration depth and incident photon energies for the uranium has been studied. It has been concluded that the values of energy absorption and exposure buildup factors are very large at 0.15 MeV.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3808-3815
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• 2021

In this paper, the EPICS2017 photoatomic database was used to evaluate the photon mass attenuation coefficients and buildup factors of soils collected at different depths in the Philippine islands. The extraction and interpolation of the library was accomplished at the recommended linear-linear scales to obtain the incoherent and total cross section and mass attenuation coefficient. The buildup factors were evaluated using the G-P fitting method in ANSI/ANS-6.4.3. An agreement was achieved between XCOM, MCNP5, and EPICS2017 for the calculated mass attenuation coefficient values. The buildup factors were reported at several penetration depths within the standard energy grid. The highest values of both buildup factor classifications were found in the energy range between 100 and 400 keV where incoherent scattering interaction probabilities are predominant, and least at the region of predominant photoionization events. The buildup factors were examined as a function of different soil silica contents. The soil samples with larger silica concentrations were found to have higher buildup factor values and hence lower shielding characteristics, while conversely, those with the least silica contents have increased shielding characteristics brought by the increased proportions of the abundant heavier oxides.

• Advances in materials Research
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• v.2 no.2
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• pp.93-98
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• 2013

Objective of this study is to calculate gamma buildup factors for pointed and isotropic gamma sources in depleted uranium, uranium dioxide, natural uranium, tin, water and concrete using MCNP4C code. The thickness of the media ranges from 0.5 to 10 mean-free-path (mfp) and gamma energy ranges from 0.5 to 10 MeV. Owing to the outstanding accuracy of MCNP in calculation involving gamma interaction, results fairly match those reported previously. The maximum relative error is 2%.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4426-4430
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• 2023

In this article, Exposure Buildup Factor(EBF) and the Energy Absorption Buildup Factor(EABF) have been determined for blood, brain, and muscle using the Monte Carlo method which is represented by MCNP5 codes and compared with geometric progression(G-P) fitting method which is represented by Phy-X/PSD online platform. The novelty of the present work is used an energy source of less than 0.1 MeV to determine buildup factors using MCNP5 and using Phy-X/PSD for some human tissues. thus, the energy range used in this case study was 0.06-3 MeV for penetration depths covered 0.5-3 MFP. Results of MCNP5 and Phy-X/PSD are validated against reference values of water that were reported at ANS-6.4.3. present results of EABFs and EBFs for the previously mentioned human tissues appeared good agreement between MCNP5 in comparison with Phy-X/PSD, whereas, the maximum average relative deviation did not exceed 2.37%. results of our article can be used in different medical applications, such as brachytherapy, radiotherapy, and diagnostics.

• Journal of Digital Convergence
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• v.3 no.1
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• pp.115-148
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• 2005

This study is intended to suggest a more reasonable and practical method of estimating discount & capitalization rate for valuation of closely-held culture content business, that is, to modify the Buildup Summation Model(which is recommended for the closely-held business by the NACVA) by adopting the weighted ratings in the CT Project Investment Evaluation of the Korea Culture Contents Association to risk factors of the Buildup Summation Model. This method is ease to apply for closely-held culture content business and has advantages in applying the weighted rates based on the characteristics of respective culture contents. And it can make up for the Weighted Average Cost of Capital (WACC) which shows generally low discount rates.

• Journal of Radiation Protection and Research
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• v.11 no.2
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• pp.123-129
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• 1986

Using Cs-137 and Co-60 gamma-ray irradiation systems, the buildup factors for 2.9 cm thick lead absorber and the air scattering factors are determined for different beam solid angles from$2.4{\pi}{\times}10^{-3}sr\;to\;17.3{\pi}{\times}10^{-3}sr$. The corresponding buildup factors are turned out to be the values from 1.054 to 1.194 and the scattering factors to be the values from 1 to 1.064, respectively. To verify our results, calculated values using these factors and experimental values are compared. The differences between them are not more than 3. 3%.

• Proceedings of the International Microelectronics And Packaging Society Conference
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• 2001.09a
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• pp.183-192
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• 2001

1- Buildup PCB technology is utilized to a bare chip attach substrate technology for packaging of semiconductor chip 2- Requirement for the substrate design rule is described in SIA International Technology Roadmap for Semiconductor. 3- There are seven fabrication methods of build-up technology. 4- Coating and lamination for resin and photo, and laser for micro via hope processes are available. Below $50\mu\textrm{m}$ in diameter is possible. 5- Fine pitch lines down to $30\mu\textrm{m}$ can be achieved by pattern plating with better electrical property. 6- Dielectric loss reduction is a key material improvement item for next generation build-up technology. 7- High band width up to 512 GB/s is possible with current wiring groundrule.

• Journal of Radiation Protection and Research
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• v.43 no.1
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• pp.20-28
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• 2018

Background: Heat energy produced in nuclear reactors and nuclear fuel cycle facilities interactions modifies the physical properties of the shielding materials containing water content. Therefore, in the present paper, effect of the heat on shielding effectiveness of the concretes is investigated for gamma and neutron. The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors. Materials and Methods: The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors of ordinary and heavy concretes were investigated using NIST data of XCOM program and Geometric Progression method. Results and Discussion: The improvement in shielding effectiveness for photon and reduction in fast neutron for ordinary concrete was observed. The change in the neutron shielding effectiveness was insignificant. Conclusion: The present investigation on interaction of gamma and neutron radiation would be very useful for assessment of shielding efficiency of the concrete used in high temperature applications such as reactors.

• 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.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.