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  • SMR full-power scaling analysis and numerical simulation for nuclear hybrid energy system testing

    Creator:
    Hoover, Kyle J.
    Abstract:
    A scaling analysis of the Oregon State University (OSU) - Multi-Application Small Light Water Reactor (MASLWR) system has been developed, and an existing RELAP5-3D model of the facility was improved. The purpose of this study is to support the research of hybrid energy systems, being led by Idaho National Laboratory...
    Resource Type:
    Masters Thesis
    Full Text:
    degree of Master of Science in Nuclear Engineering presented on December 14, 2016. Title: SMR Full

  • Comparative analysis of the Žukauskas method and data from the OSU MASLWR test facility steam generator

    Creator:
    Wuth, Bradyn J.
    Abstract:
    Some of the next generation of small modular reactors have been designed with helically coiled steam generators integrated into the reactor vessel to enhance natural circulation. According to the developers of TASS/SMR, a one-dimensional thermal hydraulic code written to model the Korean Atomic Energy Research Institute's SMART reactor, the Žukauskas...
    Resource Type:
    Masters Thesis
    Full Text:
    SG is composed of 14 tubes that enter the vessel and then coil helically around the top portion of

  • Feed-and-bleed transient analysis of OSU APEX facility using the modern Code Scaling, Applicability, and Uncertainty method

    Creator:
    Hallee, Brian Todd
    Abstract:
    The nuclear industry has long relied upon bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis accidents. These methods have been known to return highly-conservative results, limiting the operating conditions of the reactor. The Best-Estimate Plus Uncertainty (BEPU) method using a modernized version of the Code-Scaling,...
    Resource Type:
    Masters Thesis
    Full Text:
    . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.5 Forward Uncertainty Propagation . . . . . . . . . . . . . . . . . . . . . . . . 14 2.5.1

  • Liquid entrainment at an upward oriented vertical branch line from a horizontal pipe

    Creator:
    Welter, Kent B.
    Abstract:
    Under simulated accident conditions, tees in the primary coolant loop of a Pressurized Water Reactor (PWR) can deviate from their original design purpose and become separators that effectively remove core heat sink capacity. This method of primary coolant removal is a phenomelogical subset of phase separation known as liquid entrainment,...
    Resource Type:
    Dissertation
    Full Text:
    Details of the Plexiglas pipe containing the branch inserts (KfK) ................. 14 2.6 Details of the

  • Effects of fluid axial conduction on liquid metal natural circulation and linear stability

    Creator:
    Sabharwall, Piyush
    Abstract:
    The effect of fluid axial thermal conduction on one-dimensional liquid metal natural circulation and its linear stability was performed through non-dimensional analysis, steady-state assessment, and linear perturbation evaluation. The Nyquist criterion and a root search method were employed to find the linear stability boundary of both forward and backward circulations....
    Resource Type:
    Masters Thesis
    Full Text:
    ) .............................................................. 44 13. 3 D Surface Plots ofF(w) ......................................................... 47 14. Real

  • Gravity-based non-invasive superconducting gravimeter system for reactor coolant inventory monitoring

    Creator:
    Danielson, Jordan J.
    Abstract:
    All nuclear reactors rely heavily on the use of level sensing instrumentation to detect the behaviour of liquid in vessels. The majority of level sensing instruments are differential pressure sensing meters, which are invasive and could be the failure source in the pressure boundary of the primary coolant system. In...
    Resource Type:
    Masters Thesis
    Full Text:
    Pressure Transmitter model 3051S, [3] . . . 13 2.2 SG Sphere, Capacitive Plates, and Coil Configuration [4

  • Heat Loss Characterization of NIST-1's Containment Vessel

    Creator:
    Yoshiura, Ramon K.
    Abstract:
    NuScale Power has designed a small modular reactor using natural circulation cooling dependent systems for both steady-state and transient operations. The NuScale Integral Test System (NIST-1) facility was built to supply test data to NuScale Power to verify and validate computer codes for the NuScale Power Mod- ule (NPM) with...
    Resource Type:
    Masters Thesis
    Full Text:
    Generator (SG) system flow rate [5] . . . . . . . . . . . . . . 18 2.8 Types of Condensation. TS is the

  • Flow Dynamics and Condensation of Film Flows in Small Modular Reactors

    Creator:
    Lee, Dongyoung
    Abstract:
    There is renewed interest in the reliability and safety of nuclear power plants following the Fukushima Daiichi nuclear accident followed by 8.9 magnitude earthquake and Tsunami with the height of 15 m on March 11, 2011. Small Modular Reactors (SMRs) have been developed to improve safety systems by utilizing passive...
    Resource Type:
    Dissertation
    Full Text:
    ..…………………………………….....14 2.4 Major technical parameters for IRIS…...………………………………………..17 2.5 Major technical

  • Feasibility Study on MHD Energy Conversion for Applications in Liquid Metal Cooled Nuclear Reactors

    Creator:
    Qiu, Hanrui
    Abstract:
    The application of two-phase magnetohydrodynamic(MHD) in liquid metal cooled nuclear reactors presents opportunities to simplify the power generation system by eliminating moving components like a turbine for enhanced safety. The proposed system in this thesis consists of two loops with immiscible working fluids. The compressed gas expands in the two-phase...
    Resource Type:
    Masters Thesis
    Full Text:
    heat exchanger section, which transfers energy from the primary loop to the steam generator(SG). The

  • Quantified PIRT and uncertainty quantification for computer code validation

    Creator:
    Luo, Hu
    Abstract:
    This study is intended to investigate and propose a systematic method for uncertainty quantification for the computer code validation application. Uncertainty quantification has gained more and more attentions in recent years. U.S. Nuclear Regulatory Commission (NRC) requires the use of realistic best estimate (BE) computer code to follow the rigorous...
    Resource Type:
    Dissertation
    Full Text:
    coef.) for pressure in vessel.................. 104 Figure 4-14 Mean simple correlation (Pearson coef