• Nuclear Engineering and Technology
• /
• 제48권1호
• /
• pp.33-42
• /
• 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
• /
• 제42권5호
• /
• pp.520-545
• /
• 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
• /
• 제45권6호
• /
• pp.582-590
• /
• 2007

Water gas shift reaction(WGSR)이 일어나는 파이럿 규모 고온반응기에서의 거동 및 성능을 예측하기 위하여 수학적 모델을 수립하고 모사를 수행하였다. 반응기의 형상, 유체 및 열 이동에 대해 상세한 모델링이 가능한 전산유체역학 기법과 공정시스템 공학에서 사용되는 공정모사 기법을 함께 사용한 multiscale 모델링 및 모사를 수행하였으며, 그 결과를 일반 공정모사와 비교하였다. Multiscale 모사를 통해 CO의 전환율은 최고 0.85, 발열반응으로 인해 충전층의 온도는 약 720 K까지 오름을 알 수 있었다. 또한 동적모사를 통해 시간에 따른 반응기내에서의 온도분포, 전환율 분포 등의 주요한 변수 및 성능들의 시간에 따른 변화를 예측할 수 있었다. Multiscale 모사 기법은 파이럿 규모의 반응기뿐 아니라 상업규모의 공정에 대해 실제 상황을 상세히 반영하여 정확한 예측이 가능하므로, 상업공정 설계에 주요한 기술로 사용될 수 있다.

• 한국진공학회:학술대회논문집
• /
• 한국진공학회 2008년도 제34회 동계학술대회 초록집
• /
• pp.181-181
• /
• 2008

• Nuclear Engineering and Technology
• /
• 제50권4호
• /
• pp.542-552
• /
• 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
• /
• 제어로봇시스템학회 1996년도 한국자동제어학술회의논문집(국내학술편); 포항공과대학교, 포항; 24-26 Oct. 1996
• /
• pp.788-791
• /
• 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
• /
• 한국전산유체공학회 2008년도 춘계학술대회논문집
• /
• pp.268-271
• /
• 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
• /
• 제51권4호
• /
• pp.968-976
• /
• 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.

• 방사성폐기물학회지
• /
• 제22권1호
• /
• pp.37-44
• /
• 2024

In 2017, a decision was made to permanently shut down Kori Unit 1, and preparations began to be made for its decontamination and decommissioning. The dismantling of the biological shields concrete, reactor vessel (RV), and reactor vessel internals (RVI) is crucial to the nuclear decommissioning process. These components were radiologically activated by the neutron activation reaction occurring in the reactor during its operational period. Because of the radioactivity of the RV and RVI of Kori Unit 1, remotely controlled systems were developed for cutting within the cavity to reduce radiation exposure. Specialized equipment was developed for underwater cutting operations. This paper focuses on modeling related to RVI operations using the MAVRIC code and the dose calculation for a diver entering the cavity. The upper and lower parts of the RVI are classified as low-level radioactive waste, while the sides that came into contact with the fuel are classified as intermediate-level radioactive waste. Therefore, the modeling presented in this paper only considers the RVI sides because the upper and lower parts have a minimal impact on the radiation exposure. These research findings are anticipated to contribute to enhancing the efficiency and safety of nuclear reactor decommissioning operations.

• Nuclear Engineering and Technology
• /
• 제44권2호
• /
• pp.103-122
• /
• 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.