• Nuclear Engineering and Technology
• /
• v.41 no.3
• /
• pp.279-286
• /
• 2009

Improving over a previous study [1], this paper provides a Monte Carlo method for the heat conduction analysis of problems with complicated geometry (such as a pebble with dispersed fuel particles). The method is based on the theoretical results of asymptotic analysis of neutron transport equation. The improved method uses an appropriate boundary layer correction (with extrapolation thickness) and a scaling factor, rendering the problem more diffusive and thus obtaining a heat conduction solution. Monte Carlo results are obtained for the randomly distributed fuel particles of a pebble, providing realistic temperature distributions (showing the kernel and graphite-matrix temperatures distinctly). The volumetric analytic solution commonly used in the literature is shown to predict lower temperatures than those of the Monte Carlo results provided in this paper.

• Nuclear Engineering and Technology
• /
• v.55 no.1
• /
• pp.324-329
• /
• 2023

Different types of deterministic solution methods were used to solve neutron transport equations corresponding to half-space and slab albedo problems. In these types of solution methods, in addition to the error of the numerical solutions, the obtained results contain truncation and discretization errors. In the present work, a non-analog Monte Carlo method is provided to simulate the half-space and slab albedo problems with Inönü and Anlı-Güngör strongly anisotropic scattering functions. For each scattering function, the sampling method of the direction of the scattered neutrons is presented. The effects of different beams with different angular dependencies and the effects of different scattering parameters on the reflection probability are investigated using the developed Monte Carlo method. The validity of the Monte Carlo method is also confirmed through the comparison with the published data.

• Journal of Radiation Protection and Research
• /
• v.10 no.1
• /
• pp.41-49
• /
• 1985

Neutron Streaming analysis in 1300 MWe pressurized water reactor cavity was performed. In this calculation, the discrete ordinates transport codes, ANISN and DOT 3.5, and the Monte Carlo code, TRIPOLI-02 were used with the coupling code, DOTTRI. In this study IBM 3033 type computer was used. The calculated neutron fluxes and dose rates were compared with the measured data in a 900MWe pressurized water reactor cavity to show a good agreement, although some deviations in the results for each energy group were noticed. These results will be applied in the radiation shielding design of high capacity nuclear power reactors and, to the means of radiation protection in case of the reactor maintenance and the access of the reactor cavity.

• Nuclear Engineering and Technology
• /
• v.48 no.3
• /
• pp.751-757
• /
• 2016

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

• Journal of Radiation Protection and Research
• /
• v.41 no.4
• /
• pp.344-349
• /
• 2016

Background: Carbon ion therapy has achieved satisfactory results. However, patients have a risk to get a secondary cancer. In order to estimate the risk, it is essential to understand particle transportation and nuclear reactions in the patient's body. The particle transport Monte Carlo simulation code is a useful tool to understand them. Since the code validation for heavy ion incident reactions is not enough, the experimental data of the elementary reaction processes are needed. Materials and Methods: We measured neutron production double-differential cross-sections (DDXs) on a carbon bombarded with 430 MeV/nucleon carbon beam at PH2 beam line of HIMAC facility in NIRS. Neutrons produced in the target were measured with NE213 liquid organic scintillators located at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. Results and Discussion: Neutron production double-differential cross-sections for carbon bombarded with 430 MeV/nucleon carbon ions were measured by the time-of-flight method with NE213 liquid organic scintillators at six angles of 15, 30, 45, 60, 75, and $90^{\circ}$. The cross sections were obtained from 1 MeV to several hundred MeV. The experimental data were compared with calculated results obtained by Monte Carlo simulation codes PHITS, Geant4, and FLUKA. Conclusion: PHITS was able to reproduce neutron production for elementary processes of carbon-carbon reaction precisely the best of three codes.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1817-1825
• /
• 2020

Background: Cosmic ray-induced particles can lead to failure of semiconductors packaged for export during air transport. This work performed MCNP 6.2 simulations to optimize shielding against neutrons and protons induced by cosmic radiation Methods and materials: The energy spectra of protons and neutrons by incident angle at the flight altitude were determined using atmospheric cuboid model. Various candidates for the shielding materials and the geometry of the Unit Load Device Container were evaluated to determine the conditions that allow optimal shielding at all sides of the container. Results: It was found that neutrons and protons, at the flight altitude, generally travel with a downward trajectory especially for the particles with high energy. This indicated that the largest number of particles struck the top of the container. Furthermore, the simulation results showed that, among the materials tested, borated polyethylene and stainless steel were the most optimal shielding materials. The optimal shielding structure was also determined with the weight limit of the container in consideration. Conclusions: Under the determined optimal shielding conditions, a significantly reduced number of neutrons and protons reach the contents inside the container, which ultimately reduces the possibility of semiconductor failure during air transport.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3073-3084
• /
• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Nuclear Engineering and Technology
• /
• v.54 no.10
• /
• pp.3803-3810
• /
• 2022

OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. The Weight Window Mesh (WWM) function and an automatic Global Variance Reduction (GVR) method was recently developed and implemented in a developmental branch of OpenMC. This WWM function and GVR method broaden OpenMC's usage in general purposes deep penetration shielding calculations. However, the Local Variance Reduction (LVR) method, which suits the source-detector problem, is still missing in OpenMC. In this work, the Weight Window Generator (WWG) function has been developed and benchmarked for the same branch. This WWG function allows OpenMC to generate the WWM for the source-detector problem on its own. Single-material cases with varying shielding and sources were used to benchmark the WWG function and investigate how to set up the particle histories utilized in WWG-run and WWM-run. Results show that there is a maximum improvement of WWM generated by WWG. Based on the above results, instructions on determining the particle histories utilized in WWG-run and WWM-run for optimal computation efficiency are given and tested with a few multi-material cases. These benchmarks demonstrate the ability of the OpenMC WWG function and the above instructions for the source-detector problem. This developmental branch will be released and merged into the main distribution in the future.

• Nuclear Engineering and Technology
• /
• v.56 no.2
• /
• pp.673-684
• /
• 2024

STREAM, developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST), is a deterministic neutron- and photon-transport code primarily designed for light water reactor (LWR) analysis. Initially, the photon module in STREAM did not account for fluorescence and bremsstrahlung photons. This article presents recent developments regarding the integration of atomic relaxation and bremsstrahlung models into the existing photon module, thus allowing for the transport of secondary photons. The photon flux and photon heating computed with the newly incorporated models is compared to results obtained with the Monte Carlo code MCS. The incorporation of secondary photons has substantially improved the accuracy of photon flux calculations, particularly in scenarios involving strong gamma emitters. However, it is essential to note that despite the consideration of secondary photon sources, there is no noticeable improvement in the photon heating for LWR problems when compared to the photon heating obtained with the previous version of STREAM.

• Nuclear Engineering and Technology
• /
• v.49 no.6
• /
• pp.1326-1338
• /
• 2017

The continuous energy Monte Carlo neutron transport code, MC21, was coupled to the CTF subchannel thermal-hydraulics code using a combination of Consortium for Advanced Simulation of Light Water Reactors (CASL) tools and in-house Python scripts. An MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 6 demonstrated good agreement with MC21/COBRA-IE and VERA solutions. The MC21/CTF solution for VERA Core Physics Benchmark Progression Problem 7, Watts Bar Unit 1 at beginning of cycle hot full power equilibrium xenon conditions, is the first published coupled Monte Carlo neutronics/subchannel T-H solution for this problem. MC21/CTF predicted a critical boron concentration of 854.5 ppm, yielding a critical eigenvalue of $0.99994{\pm}6.8E-6$ (95% confidence interval). Excellent agreement with a VERA solution of Problem 7 was also demonstrated for integral and local power and temperature parameters.