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  • 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:
    . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 5.3 Discretized Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 5.3.1 End

  • 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:
    . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 TABLE OF CONTENTS (Continued) Page 5.3.1 Primary Loop Comparisons

  • 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:
    ] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 4.2 Detail of CVS/PRHR HX flow connnection to SG-2 lower channel head. [1] 43 4.3 Upper and

  • 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:
    ) ............................... 95 4.5 Step phenomenon recorded by ring probes (1 Hz, SG = Steam generator) ... 95 4.6 Photo and

  • 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:
    . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 6.4 Simulation Setup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 6.4.1 Physical

  • 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:
    Heat loss P Perturbation; Particular solution Core Reactor core SG Heat exchanger s Flow directional

  • 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:
    ................................................60 5.2 Inverse Heat Flux estimation : Least Squares Method .....................................63

  • 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

  • Dynamic System Scaling Application to Accelerated Nuclear Fuel Testing

    Creator:
    Yoshiura, Ramon K.
    Abstract:
    The development of nuclear fuel and materials requires is a continuous effort to investigate the acute and prolonged effects of irradiation, thermal-material stress, chemical change, or other conceivable damage mechanics acquired during normal operation or accident scenarios throughout its lifetime. As in-core fuel property measurement techniques advance to support in...
    Resource Type:
    Dissertation
    Full Text:
    Phenomenon Identi�cation Ranking Table. RIA Reactivity Initiated Accident. RR Research Reactors. SG Steam