• 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.42 no.5
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• pp.520-545
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• 2010

This paper presents the neutronics modeling capabilities of the fast reactor simulation system SHARP, which ANL is developing as part of the U.S. DOE's NEAMS program. We discuss the three transport solvers (PN2ND, SN2ND, and MOCFE) implemented in the UNIC code along with the multigroup cross section generation code $MC^2$-3. We describe the solution methods and modeling capabilities, and discuss the improvement needs for each solver, focusing on massively parallel computation. We present the performance test results against various benchmark problems and ZPR-6 and ZPPR critical experiments. We also discuss weak and strong scalability results for the SN2ND solver on the ZPR-6 critical assembly benchmarks.

• Korean Chemical Engineering Research
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• v.45 no.6
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• pp.582-590
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• 2007

In view of the analysis of the phenomena and the prediction of the performance, mathematical modelling and simulation of a high temperature pilot reactor for water gas shift reaction (WGSR) has been carried out. Multiscale simulation incorporated computational fluid dynamics (CFD) technique, which has the capability to deal with the reactor shape, fluid and energy transport with extensive degree of accuracy, and process modeling technique, which, in turn is responsible for reaction kinetics and mass transport. This research employed multiscale simulation and the results were compared with those from process simulation. From multiscale simulation, the maximum conversion of was predicted approximately 0.85 and the maximum temperature at the reactor was calculated 720 K, resulting from the heat of reaction. Dynamic simulation was also performed for the time transient profile of temperature, conversion, etc. Considering the results, it is concluded that multiscale simulation is a safe and accurate technique to predict reactor behaviors, and consequently will be available for the design of commercial size chemical reactors as well as other commercial unit operations.

• Proceedings of the Korean Vacuum Society Conference
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• 2008.02a
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• pp.181-181
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• 2008

• Nuclear Engineering and Technology
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• v.50 no.4
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• pp.542-552
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• 2018

Operators face challenges to plan alternative countermeasures when no procedure exists to address the current plant state. A model-based approach is desired to aid operators in acquiring plant resources and deriving response plans. Multilevel flow modeling (MFM) is a functional modeling methodology that can represent intentional knowledge about systems, which is essential in response planning. This article investigates the capabilities of MFM to plan alternatives. It is concluded that MFM has a knowledge capability to represent alternative means that are designed for given ends and a reasoning capability to identify alternative functions that can causally influence the goal achievement. The second capability can be applied to find originally unassociated means to achieve a goal. This is vital in a situation where all designed means have failed. A technique of procedure synthesis can be used to express identified alternatives as a series of operations. A case of station blackout occurring at the boiling water reactor is described. An MFM model of a boiling water reactor is built according to the analysis of goals and functions. The accident situations are defined by the model, and several alternative countermeasures in terms of operating procedures are generated to achieve the goal of core cooling.

• 제어로봇시스템학회:학술대회논문집
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• 1996.10b
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• pp.788-791
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• 1996

A mathematical model is developed for thermal solution copolymerization of styrene and acrylonitrile in a continuous stirred tank reactor(CSTR). Computational studies are carried out with the continuous copolymerization system model developed in this work to give the monomer conversion, copolymer composition and the average molecular weights of the copolymer. By performing the dynamic analysis of the reaction system, the polymer properties against the changes in the operating conditions are determined quantitatively. The cascade PID and fuzzy controller show satisfactory performances for both set point tracking and disturbance rejection. Especially, the fuzzy controller is superior to the PID controller.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.268-271
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• 2008

If a reactor vessel is failed to retain a molten corium in a postulated severe accident, the molten corium is released outside the reactor vessel into a reactor cavity. The molten corium would attack the concrete wall and basemat of the reactor cavity, which may lead to inevitable concrete decompositions and possible radiological releases. In the OECD/MCCI project, a series of tests were performed to secure the data for cooling the molten corium spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction). Also, a MCCI (Molten Core Concrete Interaction) analysis code, CORQUENCH was upgraded at Argonne National Laboratory with embedding the new models developed for the tests. This paper deals with analyses of MCCI at plant scale under the conditions of top flooding using the upgraded CORQUENCH code. The modeling approach is briefly summarized first, followed by presentation of a validation calculation that illustrates the predicative capability of the modeling tool. With this background in place, the model is then used to carry out a parametric set of scoping calculations that define approximate coolability envelopes for the LCS (Limestone Common Sand) concrete that has been evaluated in the OECD/MCCI project.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.968-976
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• 2019

This paper presents a number of verification case studies for a recently developed sensitivity/uncertainty code package. The code package, ROMUSE (Reduced Order Modeling based Uncertainty/Sensitivity Estimator) is an effort to provide an analysis tool to be used in conjunction with reactor core simulators, in particular the Virtual Environment for Reactor Applications (VERA) core simulator. ROMUSE has been written in C++ and is currently capable of performing various types of parameter perturbations and associated sensitivity analysis, uncertainty quantification, surrogate model construction and subspace analysis. The current version 2.0 has the capability to interface with the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) code, which gives ROMUSE access to the various algorithms implemented within DAKOTA, most importantly model calibration. The verification study is performed via two basic problems and two reactor physics models. The first problem is used to verify the ROMUSE single physics gradient-based range finding algorithm capability using an abstract quadratic model. The second problem is the Brusselator problem, which is a coupled problem representative of multi-physics problems. This problem is used to test the capability of constructing surrogates via ROMUSE-DAKOTA. Finally, light water reactor pin cell and sodium-cooled fast reactor fuel assembly problems are simulated via SCALE 6.1 to test ROMUSE capability for uncertainty quantification and sensitivity analysis purposes.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.103-122
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• 2012

Significant advances in computational performance have occurred over the past two decades, achieved not only by the introduction of more powerful processors but the incorporation of parallelism in computer hardware at all levels. Simultaneous with these hardware and associated system software advances have been advances in modeling physical phenomena and the numerical algorithms to allow their usage in simulation. This paper presents a review of the advances in computer performance, discusses the modeling and simulation capabilities required to address the multi-physics and multi-scale phenomena applicable to a nuclear reactor core simulator, and present examples of relevant physics simulation codes' performances on high performance computers.

• Transactions of the Korean hydrogen and new energy society
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• v.19 no.5
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• pp.460-466
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• 2008

In this study, a computer modeling work has been performed for the power generation system using polymer electrolyte fuel cell with Aspen Plus general purpose chemical process simulator. Stoichiometric reactor module was used for the modeling of reformer for the production of hydrogen. For the modeling of the electrochemical reaction, Gibbs reactor module built-in Aspen Plus was utilized. SRK equation of state model was selected for the proper simulation of the overall fuel cell system.