• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.291-298
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• 2014

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor ($k_{eff}$), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.870-876
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• 1995

In this paper, a new three-dimensional nodal diffusion code which is based on the AFEN methodology is described and tested. The method expands the homogeneous flux within a node in ter-ms of eighteen analytic basis functions satisfying the diffusion equation at any point of the node. And the nodal coupling equations are derived such that nodal balance, current continuity and leakage balance within an infinitesimally small box around the edge are satisfied. To verify its accuracy, the code was applied to the well-known static LMW benchmark problem and a small core benchmark problem that has the same material properties as the three-dimensional IAEA benchmark problem and compared with two other codes (QUANDRY, VENTURE). The results show that the code provides good accuracy both in the power distribution and in the effective multiplication factor.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.33-42
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• 2016

Spacer grids play an important role in maintaining the proper form of the fuel assembly structure and ensuring the safety of reactor core design. This study applies the Monte Carlo method to the analysis of the neutronics effects of spacer grids in a typical pressurized water reactor (PWR). The core problem used to analyze the neutronics effects of spacer grids is a modified version of Korea Advanced Institute of Science and Technology benchmark problem 1B, based on an Advanced Power Reactor 1400 (APR1400) core model. The spacer grids are modeled and added to this test problem in various ways. Then, by running MCNP5 for all cases of spacer grid modeling, some important numerical results, such as the effective multiplication factor, the spatial distributions of neutron flux, and its energy spectrum are obtained. The numerical results of each case of spacer grid modeling are analyzed and compared to assess which type has more advantages in accuracy of numerical results and effectiveness in terms of geometry building. The conclusion is that the most realistic modeling for Monte Carlo calculation is the "volume-preserving" streamlined heterogeneous spacer grids, but the "banded" dissolution spacer grids modeling is a more practical yet accurate model for routine (deterministic) analysis.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1259-1268
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• 2017

The alpha eigenvalue problem in multigroup neutron diffusion is studied with particular attention to the theoretical analysis of the model. Contrary to previous literature results, the existence of eigenvalue and eigenflux clustering is investigated here without the simplification of a unique fissile isotope or a single emission spectrum. A discussion about the negative decay constants of the neutron precursors concentrations as potential eigenvalues is provided. An in-hour equation is derived by a perturbation approach recurring to the steady state adjoint and direct eigenvalue problems of the effective multiplication factor and is used to suggest proper detection criteria of flux clustering. In spite of the prior work, the in-hour equation results give a necessary and sufficient condition for the existence of the eigenvalue-eigenvector pair. A simplified asymptotic analysis is used to predict bands of accumulation of eigenvalues close to the negative decay constants of the precursors concentrations. The resolution of the problem in one-dimensional heterogeneous problems shows numerical evidence of the predicted clustering occurrences and also confirms previous theoretical analysis and numerical results.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.764-769
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• 2022

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1097-1104
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• 2023

A preliminary criticality analysis for novel pyrochemical apparatuses for the reprocessing of mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor was performed. High-temperature processing apparatuses, "metallization" electrolyzer, refinery remelting apparatus, refining electrolyzer, and "soft" chlorination apparatus are considered in this work. Computational models of apparatuses for two neutron radiation transport codes (MCU-FR and MCNP) were developed and calculations for criticality were completed using the Monte Carlo method. The criticality analysis was performed for different loads of fissile material into the apparatuses including overloading conditions. Various emergency situations were considered, in particular, those associated with water ingress into the chamber of the refinery remelting apparatus. It was revealed that for all the considered computational models nuclear safety rules are satisfied.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1881-1895
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• 2020

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

• Korean Institute of Interior Design Journal
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• v.24 no.6
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• pp.229-239
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• 2015

The 21st century is a multiplication age and social and cultural phenomena have become diverse and peoples' desires and individuality have become important. Accordingly, the sensibility that reflects human taste is also required in the exhibition space. The exhibitions in this age induce the direct cognition of senses or take interactive forms that contact diverse media and react. The purpose of this research is to define the concept of haptic presentation method in which the audience perceive in the exhibition space by themselves and the visual elements spread into other senses and perceive complexly, and to present the directional nature. To conduct this research, first, this researcher recognized that haptic sensory experiential research by analyzing the roles and transition history of exhibition space is needed for the present age Second, based on philosophical theories, four haptic sensory expression characteristics (medium nature, experiential nature, attractiveness, sensitiveness) were derived by substituting Giles Deleuze's four haptic spatial characteristics (grasping short distance, dispersed gaze, cognition of bodily movement, formation of synesthesia through complex senses) and six formative factors of exhibition space (space, form, size, light, quality of materials, and color). And the effective exhibition presentation methods were analyzed through six cases of experiential exhibition spaces. Accordingly, what matters in the experiential exhibition space is to produce the four characteristics: medium nature, experientiality, attractiveness, and sensitiveness in equilibrium. It is necessary for the designers to reflect it appropriately in producing so that the audience can think and experience by themselves. Accordingly, in this thesis, it could be seen that to produce the haptic production characteristics in the experiential exhibition space in equilibrium is the important factor in the experiential exhibition space. In conclusion, experiences in the exhibition space should be approached with the transcendental haptic presentation method by which even the space of actually unexperienced cognition can be expanded and experienced through the metastasis and tension of various senses. Also, researches on such senses should be developed continuously, and this researcher expects that this will become a stimulant to present a new directivity.

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.196-203
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• 1982

To transport the spent fuel assemblies of Korea Nuclear Unit 1, which is a Westinghouse type two loop pressurized water reactor, it has been found that steel is the most appropriate material for the design of a shipping cask in comparison with lead and depleted uranium. The proposed shipping cask will transport nine fuel assemblies at the same time and is well within the weight limit of transportation by unrestricted rail car. The cask requires 33cm thick steel shield and 27cm thick water region to satisfy the 3 feet apart dose rate limit set forth in 10 CFR 71, and 1.27cm thick steel boron fuel basket to hold the fuel elements inside the cask and control the effective multiplication factor. As a safety analysis, the fuel cladding and centerline temperatures were calculated under the accident condition of complete loss of water coolant, and it was found that the temperature was much lower than the limit of the melting point. k$_{eff}$ was calculated with fresh fuel assemblies, which was found to be well lower than 0.95. For shielding computation, the multipurpose Monte Carlo code MORSE-CG and one dimensional discrete ordinates transport code ANISN were used, and the Monte Carlo codes KENO and MORSE-CG were used for criticality calculation. The radiation source terms were calculated using ORIGEN-79.9.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.47-59
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• 1993

The criticality of the shipping cask(KSC-7) for transportion of 7PWR spent fuel assemblies has been calculated and analysised on the basis of neutron transport theory. For criticality analysis, effects of the rod pitches, the fixed neutron absorbers(borated sus+boral) were considered. The effective multiplication factor has been calculated by KENO-Va, Mote Carlo method computer code, with the HANSEN-ROACH 16 group cross section set, which was made for personal computer system. The criticality for the KSC-7 cask was calculated in terms of the fresh fuel which was conservative for the aspects of nuclear critility. From the results of criticality analysis, the calculated Keff is proved to be lower than subcritical limit during normal transportation and under hypothetical accident condition. The maximum calculated criticalities of the KSC-7 were lower the safety criticality limit 1.0 recommended by US 10CFR71 both under normal and hypothetical accident condition. Also, to verify the KSC-7 criticality calculation results by using KENO-Va, it was carried out benchmark calculation with experimental data of B & W(Bobcock and Wilcox) company. From the 3s series of calculation of the KSC-7 cask and benchmark calculation, the cask was safely designed in nuclear criticality, respectively.