• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.99-106
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• 2013

Thermal fatigue is a significant long-term degradation mechanism in nuclear power plants. In particular, as operating plants become older and life time extension activities are initiated, operators and regulators need screening criteria to exclude risks of thermal fatigue and methods to determine significant fatigue relevance. In general, the common thermal fatigue issues are well understood and controlled by plant instrumentation at fatigue susceptible locations. However, incidents indicate that certain piping system Tee connections are susceptible to turbulent temperature mixing effects that cannot be adequately monitored by common thermocouple instrumentations. Therefore, in this study thermal fatigue evaluation of piping system Tee-connections is performed using the fluid-structure interaction (FSI) analysis. From the thermal hydraulic analysis, the temperature distributions are determined and their results are applied to the structural model of the piping system to determine the thermal stress. Using the rain-flow method the fatigue analysis is performed to generate fatigue usage factors. The procedure for improved load thermal fatigue assessment using FSI analysis shown in this study will supply valuable information for establishing a methodology on thermal fatigue.

• Journal of Aerospace System Engineering
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• v.16 no.3
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• pp.84-92
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• 2022

The electro-hydrostatic actuator (EHA) recently has been used in flight control fields for aircraft because of its benefits of minimizing oil leakage and weight, improving safety, and etc. while independently operating the hydraulic power source and eliminating complex hydraulic piping. The aircraft of which EHA is installed inside, has the thermal management issue of EHA, because of its limited cooling source as compared with the aircraft which installs the traditional central hydraulic system. So, the thermal analysis model which predicts the thermal characteristics of EHA, is required to resolve this thermal management issue. In this study, an oil circulation circuit inside the hydraulic power module comprised of hydraulic pump and electrical motor for EHA was applied. This is for the purpose of developing the internal rotary group of hydraulic power module, which operates under the conditions of high rotation speed and hydraulic pressure. After formulating an appropriate thermal analysis model, the thermal analysis results with oil cooled or no oil cooled hydraulic control module were compared and reviewed, for the purpose of predicting the thermal characteristics of EHA.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1231-1240
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• 2019

Unlike land-based nuclear power plants, a marine or floating reactor is affected by external forces due to ocean conditions. These external forces can cause additional accelerations and affect each system and equipment of the marine reactor. Therefore, in designing a marine reactor and evaluating its performance and stability, a thermal hydraulic safety analysis code is necessary to consider the thermal hydrodynamic effects of ship motion. MARS, which is a reactor system analysis code, includes a dynamic motion model that can simulate the thermal-hydraulic phenomena under three-dimensional motion by calculating the body force term included in the momentum equation. In this study, it was verified that the dynamic motion model can simulate fluid motion with reasonable accuracy using conceptual problems. In addition, two modifications were made to the dynamic motion model; first, a user-supplied table to simulate a realistic ship motion was implemented, and second, the flow regime map determination algorithm was improved by calculating the volume inclination information at every time step if the dynamic motion model was activated. With these modifications, MARS could simulate the thermal-hydraulic phenomena under ocean motion more realistically.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Journal of Energy Engineering
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• v.13 no.2
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• pp.101-111
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• 2004

The main purpose of a liquid metal reactor core thermal-hydraulic design is to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power distribution in the core. The thermal-hydraulic design procedure consists of the coolant flow distribution to the sub-assemblies, the coolant/fuel temperature calculations and detailed subchannel analysis. This paper describes the LMR core thermal-hydraulic design methodology and summarizes the major design and analysis results of KALIMER breeder and breakeven cores and subassemblies. KALIMER is a 150 MWe rated (392 MWth) heterogeneous core with U-TRU-Zr ternary alloy fuel and sodium coolant.

• Journal of the Korea Society for Simulation
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• v.23 no.2
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• pp.17-24
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• 2014

This study attempts to show an example of developing and applying thermal hydraulic simulation model for Aircraft-EHA. The overview of research procedure is as in the following. First, The unit hydraulic simulation model, which reflects physical quantity answering engineer's purpose is developed. Second, The unit hydraulic simulation model is combined, and then branched out to EHA hydraulic model. Third, a simulation model including flow thermal is developed, and then oil temperature rise time according to 'initial temperature and load' is examined. Finally, the master graph that can be used for designing EHA combined with thermal hydraulic analysis results in several cases is compiled, and suggested. AMESim, commercial software, is used through whole procedure.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.753-764
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• 2009

JAEA started the LSTF experiments in 1985 for the fourth stage of the ROSA Program (ROSA-IV) for the LWR thermal-hydraulic safety research to identify and investigate the thermal-hydraulic phenomena and to confirm the effectiveness of ECCS during small-break LOCAs and operational transients. The LSTF experiments are underway for the ROSA-V Program and the OECD/NEA ROSA Project that intends to resolve issues in thermal-hydraulic analyses relevant to LWR safety. Six types of the LSTF experiments have been done for both the system integral and separate-effect experiments among international members from 14 countries. Results of four experiments for the ROSA Project are briefly presented with analysis by a best-estimate (BE) code and a computational fluid dynamics (CFD) code to illustrate the capability of the LSTF and codes to simulate the thermal-hydraulic phenomena that may appear during SBLOCAs and transients. The thermal-hydraulic phenomena dealt with are coolant mixing and temperature stratification, water hammer up to high system pressure, natural circulation under high core power condition, and non-condensable gas effect during asymmetric SG depressurization as an AM action.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.561-586
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• 2012

This paper describes a summary of final prediction results by system-scale safety analysis codes during the OECD/NEA/CSNI ISP-50 exercise, targeting a 50% Direct Vessel Injection (DVI) line break integral effect test performed with the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). This ISP-50 exercise has been performed in two consecutive phases: "blind" and "open" phases. Quantitative comparisons were performed using the Fast Fourier Transform Based Method (FFTBM) to compare the overall accuracy of the collected calculations. Great user effects resulting from the combination of the possible reasons were found in the blind phase, confirming that user effect is still one of the major issues in connection with the system thermal-hydraulic code application. Open calculations showed better prediction accuracy than the blind calculations in terms of average amplitude (AA) value. A total of nineteen organizations from eleven countries participated in this ISP-50 program and eight leading thermal-hydraulic system analysis codes were used: APROS, ATHLET, CATHARE, KORSAR, MARS-KS, RELAP5/MOD3.3, TECH-M-97, and TRACE.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1388-1395
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• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.