• Structural Engineering and Mechanics
• /
• 제38권1호
• /
• pp.65-79
• /
• 2011

Concrete hydraulic structures such as: Dams, Intake Towers, Piers and dock are usually recognized as" Vital and Special Structures" that must have sufficient safety margin at critical conditions like when earthquake occurred as same as normal servicing time. Hence, to evaluate hydrodynamic pressures generated due to seismic forces and Fluid-Structure Interaction (FSI); introduction to fluid-structure domains and interaction between them are inevitable. For this purpose, first step is exact modeling of water-structure and their interaction conditions. In this paper, the basic equation involved the water-structure-foundation interaction and the effective factors are explained briefly for concrete hydraulic structure types. The finite element modeling of two concrete gravity dams with 5 m, 150 m height, reservoir water and foundation bed rock is idealized and then the effects of fluid domain and bed rock have been investigated on modal characteristic of dams. The analytical results obtained from numerical studies and modal analysis show that the accurate modeling of dam-reservoir-foundation and their interaction considerably affects the modal periods, mode shapes and modal hydrodynamic pressure distribution. The results show that the foundation bed rock modeling increases modal periods about 80%, where reservoir modeling changes modal shapes and increases the period of all modes up to 30%. Reservoir-dam-foundation interaction increases modal period from 30% to 100% for different cases.

• 한국수자원학회논문집
• /
• 제34권5호
• /
• pp.543-549
• /
• 2001

수리학적 하도추적으로부터 하천유량을 간접추정하였다. 짧은 하천구간의 세 지점 연속수위자료와 하천 단면 자료만을 사용하여 조도계수 추정b과 하천유량 계산이 가능하였다. 유량의 간접추정 과정에서, 상류-하류 경계조건을 수위-수위 조건으로 사용하였다. 미국 미시시피 강의 상류 구간 자료가 사용되었고 수리학적 하도추적에는 DWO-PER (operational dynamic wave model)를 이용하였다. DWOPER 모형에서 수정 Newton-Raphson법에 의한 조도계수 추정과정을 개선하기 위하여 SCE-UA 전역최적화 기법을 적용하였으며, SCE-UA 기법의 결과가 적은 오차를 보였다. 특정 홍수에 대하여 유량을 추정한 결과, 몇 개를 제외한 대부분의 계산유량이 10% 이내의 오차를 보였다.

• Nuclear Engineering and Technology
• /
• 제33권6호
• /
• pp.638-651
• /
• 2001

The present study is to assess the applicability of the best-estimate thermal-hydraulic code, MARS 1.4, for the analysis of thermal-hydraulic behavior in PWRs during natural circulation conditions. The code simulates a natural circulation test, BETHSY test 4. la, which was conducted on the integral test facility of BETHSY. The test represented the cooling states of the primary cooling system under single-phase natural circulation, two-phase natural circulation and the reflux condensation mode with conditions corresponding to the residual power, 2% of the rated core power value and 6.8 MPa at the secondary system. Based on MARS 1.4 calculations, the major thermal-hydraulic behaviors during natural circulation are evaluated and the differences between the experimental data and calculated results are identified. The calculated results show generally good behavior with regard to the experimental results; the region of single-phase natural circulation is 100-92% of the initial mass inventory, two-phase natural circulation is 84-63 %, and the reflux condensation mode occurred below 58 %. U-tubes empty and the core uncovery are obtained at 39 % and 34 % of the initial mass inventory, respectively.

• International Journal of Fluid Machinery and Systems
• /
• 제5권4호
• /
• pp.152-160
• /
• 2012

Model tests and CFD were carried out to find out the cause of cavitation surge in hydraulic power plants. In experiments the cavitation surge was observed at flow rate, both with and without a surge tank placed just upstream of the inlet volute. The surge frequency at smaller flow rate was much smaller than the swirl mode frequency caused by the whirl of vortex rope. An unsteady CFD was carried out with two boundary conditions: (1) the flow rate is fixed to be constant at the volute inlet, (2) the total pressure is kept constant at the volute inlet, corresponding to the experiments without/with the surge tank. The surge was observed with both boundary conditions at both higher and lower flow rates. Discussions as to the cause of the surge are made based on additional tests with an orifice at the diffuser exit, and with the diffuser replaced with a straight pipe.

• Nuclear Engineering and Technology
• /
• 제50권2호
• /
• pp.297-305
• /
• 2018

Fuel assembly (FA) bow in pressurized water reactor (PWR) cores is considered to be a complex process with a large number of influencing mechanisms and several unknowns. Uncertainty and sensitivity analyses are a common way to assess the predictability of such complex phenomena. To perform such analyses, a structural model of a row of 15 FAs in the reactor core is implemented with the finite-element code ANSYS Mechanical APDL. The distribution of lateral hydraulic forces within the core row is estimated based on a two-dimensional Computational Fluid Dynamics model with porous media, assuming symmetric or asymmetric core inlet and outlet flow profiles. The influence of the creep rate on the bow amplitude is tested based on different creep models for guide tubes and fuel rods. Different FA initial states are considered: fresh FAs or FAs with higher burnup, which may be initially straight or exhibit an initial bow from previous cycles. The simulation results over one reactor cycle demonstrate that changes in the creep rate and the hydraulic conditions may have a considerable impact on the bow amplitudes and the bow patterns. A good knowledge of the specific creep behavior and the hydraulic conditions is therefore crucial for making reliable predictions.

• 한국안전학회지
• /
• 제36권2호
• /
• pp.111-119
• /
• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Nuclear Engineering and Technology
• /
• 제54권5호
• /
• pp.1860-1873
• /
• 2022

Nuclear thermal-hydraulic system analysis codes have been developed to comprehensively model nuclear reactor systems to evaluate the safety of a nuclear reactor system. For analyzing complex systems with finite computational resources, system codes usually solve simplified fluid equations for coarsely discretized control volumes with one-dimensional assumptions and replace source terms in the governing equations with constitutive relations. Wall boiling heat transfer models are regarded as essential models in nuclear safety evaluation among many constitutive relations. The wall boiling heat transfer models of two widely used nuclear system codes, RELAP5 and TRACE, are analyzed in this study. It is first described how wall heat transfer models are composed in the two codes. By utilizing the same method described in Part 1 paper, heat fluxes from the two codes are compared under the same thermal-hydraulic conditions. The significant factors for the differences are identified as well as at which conditions the non-negligible difference occurs. Steady-state simulations with both codes are also conducted to confirm how the difference in wall heat transfer models impacts the simulation results.

• 한국콘크리트학회:학술대회논문집
• /
• 한국콘크리트학회 1999년도 학회창립 10주년 기념 1999년도 가을 학술발표회 논문집
• /
• pp.647-650
• /
• 1999

Deterioration of agricultural hydraulic structures(AHS), which are under harsh environmental conditions, is more sever than other ordinary structures. To investigate the deterioration factors of AHS, various physical and chemical analyses are performed. The porosity of AHS increases more rapidly than ordinary structures because they are subject to frequent water permeation and water-soluble materials are easily emitted to surface area. Thus, AHS are tend to be damaged by freezing and thawing more easily due to the increase of water containment inside concrete.

• 제어로봇시스템학회:학술대회논문집
• /
• 제어로봇시스템학회 1996년도 한국자동제어학술회의논문집(국내학술편); 포항공과대학교, 포항; 24-26 Oct. 1996
• /
• pp.1469-1472
• /
• 1996

Hydraulic Hot Strip Mill (HHSM) rolls materials whose size and stiffness are various. So a roll gap controller for HHSM was designed using TDC(Time Delay Control) method. The performance of the roll gap control was evaluated through computer simulations. The simulation results indicate that TDC method show excellent robustness and tracking properties against PID control method in various rolling conditions.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.193-198
• /
• 1996

In this study the RETRAN-03 code was modified to analyze the thermal-hydraulic transients under three-dimensional ship motions for the application to the future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations are successfully simulated at various conditions.