• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2712-2722
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• 2023

For an accurate and efficient time-dependent Monte Carlo (TDMC) neutron transport analysis, several advanced methods are newly developed and implemented in the Seoul National University Monte Carlo code, McCARD. For an efficient control of the neutron population, a dynamic weight window method is devised to adjust the weight bounds of the implicit capture in the time bin-by-bin TDMC simulations. A moving geometry module is developed to model a continuous insertion or withdrawal of a control rod. Especially, the history-based batch method for the TDMC calculations is developed to predict the unbiased variance of a bin-wise mean estimate. The developed methods are verified for three-dimensional problems in the C5G7-TD benchmark, showing good agreements with results from a deterministic neutron transport analysis code, nTRACER, within the statistical uncertainty bounds. In addition, the TDMC analysis capability implemented in McCARD is demonstrated to search the optimum detector positions for the pulsed-neutron-source experiments in the Kyoto University Critical Assembly and AGN201K.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.920-927
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• 2017

With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.381-390
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• 2009

This article provides selects of outstanding problems in computational neutron transport, with some suggested approaches thereto, as follows: i) ray effect in discrete ordinates method, ii) diffusion synthetic acceleration in strongly heterogeneous problems, iii) method of characteristics extension to three-dimensional geometry, iv) fission source and $k_{eff}$ convergence in Monte Carlo, v) depletion in Monte Carlo, vi) nuclear data evaluation, and vii) uncertainty estimation, including covariance data.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1211-1218
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• 2017

Monte Carlo neutron transport codes generally use the method of successive generations to converge the fission source distribution to-and then maintain it at-the fundamental mode. Recently, a phenomenon called "clustering" has been noted, which produces fission distributions that are very far from the fundamental mode. In this study, a mathematical model of clustering in Monte Carlo has been developed. The model draws on previous work for continuous-time birth-death processes, as well as methods from the field of population genetics.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.477-484
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• 2020

The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately. A modified method was applied to the subcriticality monitoring process, in which only a single control rod cluster was fully withdrawn from the core. In order to correct the error in the results obtained by Neutron Source Multiplication Method, which is based on one point reactor model, Monte Carlo neutron transport code was employed to calculate the fission neutron distribution, the iterated fission probability and the neutron flux in the neutron detector. This article analyzed the effect of a coarse mesh and a fine mesh to tally fission neutron distributions, the iterated fission probability distributions and to calculate correction factors. The subcriticality before and after modification is compared with the subcriticality calculated by MCNP code. The modified results turn out to be closer to calculation. It's feasible to implement the modified NSM method in large local reactivity addition process using Monte Carlo code based on 3D model.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.916-921
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• 2024

The control rod worth sensitivity to the neutron detector position in the rod drop experiment is studied by the time-dependent Monte Carlo (TDMC) neutron transport calculations for AGN-201K educational reactor and the Kyoto University Critical Assembly. The TDMC simulations of the rod drop experiments are conducted by the Seoul National University Monte Carlo (MC) code, McCARD, yielding time-dependent neutron densities at detector positions. The detector-position-dependent results of the total control rod worth calculated by the extrapolation, the integral counting, and the inverse methods are compared with the numerical reference using the MC eigenvalue calculations and the experimental results. From these comparisons, it is observed that the total control rod worth can be estimated with a considerable difference depending on the detector position through the rod drop experiment. The proposed TDMC simulation of the rod drop experiment can be applied for searching a better detector position or quantifying a bias for the control rod worth measurement.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1199-1210
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• 2017

A new in-line method for sampling neutrons emitted in (${\alpha}$,n) reactions based on alpha particle source information has been developed for continuous-energy Monte Carlo simulations. The new method uses a continuous-slowing-down model coupled with (${\alpha}$,n) cross section data to precompute the expected neutron yield over the alpha particle lifetime. This eliminates the complexity and computational cost associated with explicit charged particle transport. When combined with an integrated alpha particle decay source sampling capability, the proposed method provides an efficient and accurate method for sampling (${\alpha}$,n) neutrons based solely on nuclide inventories in the problem, with no additional user input required. Results from several example calculations show that the proposed method reproduces the (${\alpha}$,n) neutron yields and energy spectra from reference experiments and calculations.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1769-1780
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• 2022

A new method of dose calculation algorithm, called GPU-accelerated Monte Carlo and collapsed cone Convolution (GMCC) was developed to improve the calculation speed of BNCT treatment planning system. The GPU-accelerated Monte Carlo routine in GMCC is used to simulate the neutron transport over whole energy range and the Collapsed Cone Convolution method is to calculate the gamma dose. Other dose components due to alpha particles and protons, are calculated using the calculated neutron flux and reaction data. The mathematical principle and the algorithm architecture are introduced. The accuracy and performance of the GMCC were verified by comparing with the FLUKA results. A water phantom and a head CT voxel model were simulated. The neutron flux and the absorbed dose obtained by the GMCC were consistent well with the FLUKA results. In the case of head CT voxel model, the mean absolute percentage error for the neutron flux and the absorbed dose were 3.98% and 3.91%, respectively. The calculation speed of the absorbed dose by the GMCC was 56 times faster than the FLUKA code. It was verified that the GMCC could be a good candidate tool instead of the Monte Carlo method in the BNCT dose calculations.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3528-3542
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• 2021

A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were <3%. When studying axial power density profiles the differences in axial offset were less than 2.3% for hot full power condition. To further confirm the applicability of the developed model, the measurements with in-core neutron detectors were compared to the calculations, where differences of 5% were observed.