• Geophysics and Geophysical Exploration
• /
• v.17 no.3
• /
• pp.155-162
• /
• 2014

We performed MCNP modeling for density log, and examined its reliability and validity comparing the correction curves from physical borehole model. Based on the constructed numerical model, numerical modelings of density sonde in three-inch borehole were carried out under the various conditions such as the existence and type of casing or fluid, and also the stand-off between the sonde and borehole wall. These results of numerical modeling quantitatively reflect effects of casing and fluid in borehole, and moreover, demonstrate constant patterns with interval change from borehole wall. From this study, numerical modeling using MCNP shows a good applicability for well logging, and therefore, can be efficiently used for the calibration of well logging data under the various borehole conditions.

• The Journal of Engineering Geology
• /
• v.27 no.4
• /
• pp.429-438
• /
• 2017

Monte Carlo N-Particle (MCNP) modeling algorithm based on the Monte Carlo method was used to perform the simulation of neutron logging in order to increase the reliability and utilization of neutron logs applied in geological and resource engineering fields. To perform the simulation using MCNP, we used a realistic three-dimensional configuration of neutron sonde and formation. Validation of the modeling was confirmed by comparing the calibration curves of sonde manufacture with those calculated by MCNP modeling. After the validation, lithology effects, pore fluid effects, borehole diameter change, casing effect, and effects of borehole water level were investigated through modeling experiments. Numerical tests indicate that changes in neutron count ratio according to the lithology were quantitatively understood. In case of a borehole with a diameter of 3 inches, ratio of counting rates was higher than expected to be interpreted as borehole fluid has small effects on neutron logging. Effect of casing was also small in general, particular when porosity increases. Since modeling results above the groundwater level showed a tendency opposite to those below the groundwater level, neutron logs can be used to detect groundwater level. The modeling results simulated in this study for various borehole environments are expected to be used for data processing and interpretation of neutron log.

• Nuclear Engineering and Technology
• /
• v.52 no.4
• /
• pp.841-845
• /
• 2020

An airborne alpha beta detection system using passivated implanted planar silicon (PIPS) detector was modeled with the MCNP6 code and its resolution and detection efficiency were analyzed. Simulation of the resolution performed using the Gaussian energy broadening (GEB) function showed that the full width at half maximum (FWHM) of 35.214 keV for alpha particles was within 34-38 KeV, which is the FWHM range of the actual detector, and the FWHM of 15.1 keV for beta particles was constructed with a similar model to 17 keV, which is the FWHM range of an actual detector. In addition, the detection efficiency and the resolution were simulated according to the distance between the detector and the air filter. When the distance was decreased to 0.2 cm from 0.8 cm, the efficiency of the alpha and beta particles detection decreased from 5.33% to 4.89% and from 5.64% to 4.27%, respectively, and the FWHM of the alpha and beta particles improved from 40.9 KeV to 29.84 keV and 25.76 keV-13.27 keV, respectively.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.179-184
• /
• 1996

Power density distribution and criticality of a pressurized water reactor are calculated with a Monte Carlo calculation using the MCNP code. The MCNP model is based on one-eighth core symmetry. Individual fuel assemblies are modeled with fullscope three dimensional description except grid spacer. The fuel rod is divided into eight axial segments. Core internals above and below the active fuel region is represented as coolant. After 400 cycle calculations, the system converges to a k value of 1.09151$\pm$0.00066. Fission reaction rate in each rod is also calculated to use as the source term in pressure vessel fluence calculation.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.10a
• /
• pp.110-110
• /
• 1998

• Journal of Sensor Science and Technology
• /
• v.25 no.4
• /
• pp.264-270
• /
• 2016

Self-powered neutron detector (SPND) is a sensor to monitor a neutron flux proportional to a reactor power of the nuclear power plants. Since an SPND is usually installed in the reactor core and does not require additional outside power, it generates electrons itself from interaction between neutrons and a neutron-sensitive material called an emitter, such as rhodium and vanadium. This paper presents the simulations of the depletion sensitivity evaluations based on MCNP models of rhodium and vanadium SPNDs and light water reactor fuel assembly. The evaluations include the detail geometries of the detectors and fuel assembly, and the modeling of rhodium and vanadium emitter depletion using MCNP and ORIGEN-S codes, and the realistic energy spectrum of beta rays using BETA-S code. The results of the simulations show that the lifetime of an SPND can be prolonged by using vanadium SPND than rhodium SPND. Also, the methods presented here can be used to analyze a life-time of those SPNDs using various emitter materials.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.47-48
• /
• 2005

• Nuclear Engineering and Technology
• /
• v.41 no.5
• /
• pp.729-738
• /
• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Journal of the Korean Society of Radiology
• /
• v.16 no.5
• /
• pp.527-536
• /
• 2022

The radiation shielding characteristic of neutron shielding material has been studied as the preliminary study in order to design cosmic-ray shielding material. Specially, Soft Magnetic Material, known to be effective in EMP and radiation shielding, has been investigated to check if the material would be applicable to cosmic-ray shielding. In this work, thermal neutron shielding experiment was conducted and the Monte Carlo N-Particle(MCNP) was applied to employ skymap.dat, which is cosmic-ray data embedded in MCNP. As a result, polyethylene, borated polyethylene, and carbon nano tube, containing carbon or hydrogen, have been found to be effective in reduction of neutron flux below 20 MeV (including thermal, epithermal, evaporation). In contrast, the materials composed of iron such as SS316 and Soft Magnetic Material show a good shielding performance in the cascade energy range (above 20 MeV). Since Soft Magnetic Material is consisting of 13% of boron, it can also decrease thermal neutron flux, so it is expected that it would show a significant reduction on the entire range of neutron energy if the Soft Magnetic Material is used with hydrogen and carbon, so called low Z material.

• Journal of Radiation Protection and Research
• /
• v.37 no.4
• /
• pp.197-201
• /
• 2012

Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.