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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

  • Scaling analysis of the OSU high temperature test facility during a pressurized conduction cooldown event using RELAP5-3D

    Creator:
    Castañeda, Juan A.
    Abstract:
    In early 2000, the Generation IV International Forum (GIF) was created to perform research and development for the next generation nuclear systems. Among the selected nuclear systems was the Very High Temperature Gas-Cooled Reactor (VHTR). Then in 2008, the U.S. Department of Energy (DOE) decided that the Next Generation Nuclear...
    Resource Type:
    Masters Thesis
    Full Text:
    ............................................................................. 14 2 LITERATURE REVIEW

  • System model of a natural circulation integral test facility

    Creator:
    Galvin, Mark Robert, 1967-
    Abstract:
    The Department of Nuclear Engineering and Radiation Health Physics (NE/RHP) at Oregon State University (OSU) has been developing an innovative modular reactor plant concept since being initiated with a Department of Energy (DoE) grant in 1999. This concept, the Multi-Application Small Light Water Reactor (MASLWR), is an integral pressurized water...
    Resource Type:
    Dissertation
    Full Text:
    Table Page 2.1 OSU MASLWR Test Facility Steam Generator Coil Curvature . . . . . . . 14 3.1 Primary

  • 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

  • Multi application small light water reactor containment analysis and design

    Creator:
    Haugh, Brandon Patrick
    Abstract:
    This thesis presents the assessment of the Multi Application Small Light Water Reactor (MASLWR) containment design during steady-state and transient conditions. The MASLWR project is a joint effort between Idaho National Environmental and Engineering Laboratory (INEEL), NEXANT Bechtel, and Oregon State University. The project is funded under a Nuclear Energy...
    Resource Type:
    Masters Thesis
    Full Text:
    ............................................................................. 14 2.1.3.3 Mist

  • Thermal hydraulic analysis of a U-tube steam generator in a pressurized water reactor

    Creator:
    Chadly, Peter Brent
    Abstract:
    A vital safety concern in the analysis of nuclear power reactors is the thermal hydraulic behavior of the reactor coolant system under steady state and transient conditions. The goal of this study has been to develop an appropriate model for the U-tube type steam generators in a typical pressurized water...
    Resource Type:
    Masters Thesis
    Full Text:
    SUMMARY OF MODEL 51 STEAM GENERATOR DESIGN SPECIFICATIONS 46 13 TRANSIENT LIQUID LEVEL vs. TIME 48 14

  • Nonintrusive intelligent monitoring for nuclear power plant emergency classification

    Creator:
    Greene, Kenneth R. (Kenneth Ray), 1958-
    Abstract:
    A prototype real-time process monitoring emergency classification expert system, RT/EM-CLASS, has been developed for use at the Trojan Nuclear Power Plant. This knowledge-based system features the integration of electronically sensed plant data with the menu selection data representation of its predecessor, EM-CLASS. This prototype demonstrates the techniques required to acquire...
    Resource Type:
    Dissertation
    Full Text:
    With Module Descriptions 65 14. Plant Sensor Dynamic Model 67 15. Sample Pressure Sensor Simulated

  • A knowledge-based approach for monitoring and situation assessment at nuclear power plants

    Creator:
    Heaberlin, Joan Oylear
    Abstract:
    An approach for developing a computer-based aid to assist in monitoring and assessing nuclear power plant status during situations requiring emergency response has been developed. It is based on the representation of regulatory requirements and plant-specific systems and instrumentation in the form of hierarchical rules. Making use of inferencing techniques...
    Resource Type:
    Dissertation
    Full Text:
    indicate temperatures higher than the hot leg temperature. 14 The system also tracks time-dependent

  • 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

  • Pressurizer surge line Counter Current Flow Limitation during AP600 Mode 5 Cold Shutdown

    Creator:
    Colpo, Sarah E.
    Abstract:
    Counter Current Flow Limitation (CCFL) was observed in the pressurizer surge line of the Oregon State University APEX facility during test NRC-10. This test simulated a one-inch diameter cold leg break with a failure of three of four of the fourth-stage Automatic Depressurization System (ADS) valves. The result was a...
    Resource Type:
    Masters Thesis
    Full Text:
    Velocity for test NRC-35. 27 14. Kutateladze Number in Pressurizer Surge Line During NRC-35. 28 15

  • 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

  • An investigation of fabric composite heat pipe feasibility issues

    Creator:
    Marks, Timothy S.
    Abstract:
    The design of a fabric composite heat pipe has been completed. It is composed of two end caps, between which a fluid containment liner composed of metal foil and an outer structural layer composed of a ceramic fabric is stretched. The interior of the heat pipe is layered with a...
    Resource Type:
    Masters Thesis
    Full Text:
    . Wicking Constant. 76 A.14. Confidence Factor for Polynomial Fit. 77 A.15. Effective Pore Radius, Fibrous

  • 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

  • Classification Modeling of Nuclear Power Plant Outage Severities with Complement Naïve Bayes and Bidirectional Encoder Representations from Transformers

    Creator:
    Castillo, Michael A.
    Abstract:
    Low-level nuclear power plant outages in the United States can lead to unanticipated costs, potentially compromising the expected operation lifetime of the plant. Nuclear power plants are complex systems of interfacing components and highly regulated processes. This inherent complexity makes predicting outages from system dependencies very challenging. When outages do...
    Resource Type:
    Masters Thesis
    Full Text:
    .................................................................................................14 4 Natural Language Processing in Nuclear Power

  • A study of thermal stratification in the cold legs during the subcooled blowdown phase of a loss of coolant accident in the OSU APEX Thermal Hydraulic Testing Facility

    Creator:
    Wachs, Daniel M.
    Abstract:
    The phenomena of interest in this work is the thermal stratification which occurs during the early stages of a loss of coolant accident (LOCA) in the OSU APEX Thermal Hydraulic Test Facility, which is a scaled model of the Westinghouse AP600 nuclear power plant. Thermal stratification has been linked to...
    Resource Type:
    Masters Thesis
    Full Text:
    2-4 Prediction of mixed mean (radially averaged) temperature for axisymmetric jets (plumes). 14

  • Investigation of the IRWST flow patterns during a simulated station blackout experiment on the OSU APEX facility

    Creator:
    Strohecker, Mark F.
    Abstract:
    The OSU/APEX thermal hydraulic test facility models the passive safety systems of the Westinghouse AP600 advanced light water reactor design. Numerous experiments have been performed to test these systems, the one of focus here is the station blackout scenario. This experiment simulated the complete loss of AC power to all...
    Resource Type:
    Masters Thesis
    Full Text:
    to zero if Tl is less than Tsat. 14 The vapor generation rate as a result of sub-cooled boiling

  • RELAP5-3D modeling of ADS blowdown of MASLWR facility

    Creator:
    Bowser, Christopher Jordan
    Abstract:
    Oregon State University has hosted an International Atomic Energy Agency (IAEA) International Collaborative Standard Problem (ICSP) through testing conducted on the Multi-Application Small Light Water (MASLWR) facility. The MASLWR facility features a full-time natural circulation loop in the primary vessel and a unique pressure suppression device for accident scenarios. Automatic...
    Resource Type:
    Masters Thesis
    Full Text:
    ) 14 WF = ( 2m∑ n=0 |∆̃F (fn)|fn ) ÷ ( 2m∑ n=0 |∆̃F (fn)| ) (2.6) where ∆F (t) is the

  • Pair production treatments for deterministic pulse-height distribution simulations

    Creator:
    MacQuigg, Michael R.
    Abstract:
    This thesis presents methods for treating annihilation photon pairs in deterministic pulse height distribution (PHD) simulations. The methods are applied in PHD simulations for monoenergtic sources of 1.6 and 2.6 MeV photons incident on 5 and 10 cm 1-D slabs of germanium, sodium iodide, and lead and the results are...
    Resource Type:
    Masters Thesis
    Full Text:
    Particle Transport Equation . . . . . . . . . . . . . . . . 14 1.6.1 Slab Geometry Transport Equation

  • An Active Interrogation Technique Based on Photofission for Non-destructive Assay of Used Nuclear Fuel

    Creator:
    Wen, Xianfei
    Abstract:
    High-energy delayed γ-rays from photofission were demonstrated to be signatures for detection and identification of special nuclear materials. Such γ-rays were measured in between linac pulses using independent data acquisition systems. A list-mode system was developed to measure low-energy delayed γ-rays after irradiation. Photofission product yields of ²³⁸U and ²³⁹Pu...
    Resource Type:
    Dissertation
    Full Text:
    Page Figure 1.1 Principal γ-ray signatures of nuclear materials of interest [14

  • Development of a Novel MEMS Thermal Sensor Array for Pebble Bed HTGRs

    Creator:
    Hertel, Matthew A.
    Abstract:
    A novel high temperature thermal instrument is proposed, which utilizes a dual-sensing platinum resistor encased in a mono-crystalline alpha-alumina (sapphire) substrate. The instrument is comprised of four platinum trace elements, oriented with 90° rotational symmetry atop a 1120 oriented crystal lattice substrate. All four of these sensors measure temperature directly...
    Resource Type:
    Dissertation
    Full Text:
    . of the PBMR-CG and MHTGR designs [13] [14] [15] [16] ................ 6 Figure 1-3: Examples of

  • Applying uncertainty and sensitivity on thermal hydraulic subchannel analysis for the multi-application small light water reactor

    Creator:
    Brigantic, Adam
    Abstract:
    Small modular reactors (SMRs) are a recent advancement in commercial nuclear reactor design with growing interest worldwide. New SMR concepts, such as the Multi-Application Small Light Water Reactor (MASLWR), must undergo a licensing processes established by the U.S. Nuclear Regulatory Commission (NRC) prior to commercial operation. Given the lack of...
    Resource Type:
    Masters Thesis
    Full Text:
    ..................................................................................................................... 14 3. MASLWR DESCRIPTION

  • 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

  • A comparison of three coarse-mesh nodal methods for boiling water reactor analysis

    Creator:
    Cho, Byung-Oh
    Abstract:
    Computer codes for versions of STORM, capable of calculating power distributions in a three-dimensional quarter-core of a boiling water reactor, were developed by using both a modified one-group and two-group diffusion theory model with several different power calculation schemes. Several numerical examples that demonstrate the high efficiency of the coarse-mesh...
    Resource Type:
    Masters Thesis
    Full Text:
    Solution for Node-Center Fluxes 10 2.3 Node Average Power Calculation 14 3.0 A BORRESEN'S COARSE-MESH

  • An on-line advanced plant simulator (OLAPS)

    Creator:
    Samuels, Jeffery W.
    Abstract:
    A PC based on-line advanced plant simulator (OLAPS) for high quality simulations of Portland General Electric's Trojan Nuclear Facility is presented. OLAPS is designed to simulate the thermal-hydraulics of the primary system including core, steam generators, pumps, piping and pressurizer. The simulations are based on a five equation model that...
    Resource Type:
    Dissertation
    Full Text:
    Philosophy 9 1.2.2 Enhancement to the State of the Art 11 2. Component Models 14 2.1 Core Model 14 2.2

  • Dynamic simulation of a Westinghouse PWR four-loop plant during steam generator tube rupture event

    Creator:
    Sui, Yueh Chun
    Abstract:
    This study addresses the application of a full reactor system computer code to simulate the reactor primary and secondary loop thermal hydraulics, plant control and protection systems during the steam generator tube rupture (SGTR) events. The Trojan nuclear plant which is owned and operated by the Portland General Electric (PGE)...
    Resource Type:
    Masters Thesis
    Full Text:
    Description 14 2.4.1 Reactor Control System 14 2.4.2 Charging, Letdown and Makeup Water Control System 16

  • Generation of Multigroup Multiband Cross Sections with Monte Carlo Simulations

    Creator:
    Lam, Adam Q.
    Abstract:
    This is the investigation into The generation of high-fidelity multigroup multiband cross sections from Monte Carlo neutron transport simulations. Previous methods for generating multigroup multiband (MGMB) cross sections, and multigroup cross sections, assume an approximate shape for the scalar flux. This approximate flux shape is the product of an energy-dependent...
    Resource Type:
    Masters Thesis
    Full Text:
    sections for several isotopes. . . . . . . 14 4 MG (2-groups) cross sections vs energy

  • Probabilistic risk assessment of a vented fuel system utilized in a generation IV gas-cooled fast reactor nuclear power plant

    Creator:
    Lenhof, Renae
    Abstract:
    A limited scope Probabilistic Risk Assessment (PRA) of a vented nuclear fuel system for a Generation IV Gas-Cooled Fast Reactor Plant was performed. The goal of the study was to better understand the safety and licensing implications of vented fuel technology. A Level 1 PRA was performed to determine the...
    Resource Type:
    Masters Thesis
    Full Text:
    ) ................................................................... 14 2.2.4 Energy Multiplier Module (EM2

  • 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

  • Temperature profiles of spheres packed in regular arrays

    Creator:
    Cunningham, Mitchel Eugene
    Abstract:
    An analytical solution is developed for the temperature profile of a unit cell consisting of a sphere of one material centered in a cube of a second material. All material properties are considered constant and boundary conditions consist of specified temperatures on the top and bottom faces of the cube...
    Resource Type:
    Masters Thesis
    Full Text:
    the unit cell 12 6 Coordinate systems imposed on the unit cell 14 7 Returned accuracy versus number

  • 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:
    . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2.5 Fuel-Pellet (-Slug) Phenomena . . . . . . . . . . . . . . . . . . . . . . 14 2.5.1 Pellet

  • Design and analysis of a nuclear reactor core for innovative small light water reactors

    Creator:
    Soldatov, Alexey I.
    Abstract:
    In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design...
    Resource Type:
    Dissertation
    Full Text:
    . Effective Fraction of Delayed Neutrons (MASLWR)……………. 114 14 Figure LIST OF FIGURES (Continued

  • Quantifying System Distortions between the VTR and IVTR facilities through H2TS Scaling Methodology and RELAP5-3D Simulations

    Creator:
    Johnson, Juwan A.
    Abstract:
    Sodium Cooled Fast Reactors (SFR) are complex and expensive due to high operating temperatures and neutron flux which pose technical challenges for these reactors. To overcome these challenges with SFRs, research and development of small scale and large-scale facilities are needed to bridge the gap in understanding. Scaling analyses develop...
    Resource Type:
    Masters Thesis
    Full Text:
    ] ...................................................................... 14 Figure 2.5: The role of H2TS [2

  • Study of interfacial condensation in a nuclear reactor core makeup tank

    Creator:
    Ma, Chang Chun
    Abstract:
    Steam interfacial condensation in a core makeup tank was simulated using the code RELAP5/MOD3 version 8.0 to predict the violent pressure oscillation phenomena in a core makeup tank. Six base cases were carried out to study the effects of back pressure and of vacuum conditions produced in the core makeup...
    Resource Type:
    Masters Thesis
    Full Text:
    between the pressure in the steam supply tank and the pressure in the CMT tank. 14 The outlet

  • Characterization of fission product transport in a gen. IV gas-cooled fast reactor plant utilizing vented fuel

    Creator:
    Deason, Wesley
    Abstract:
    Fission product transport in a Gen. IV Gas-Cooled Fast Reactor Plant utilizing vented fuel has been characterized using analytical and computational methods. The goal was to increase current understanding of fission product transport in helium-cooled GFRs using vented fuel and to provide a toolset for determining issues which may arise...
    Resource Type:
    Masters Thesis
    Full Text:
    Helium Purification Systems ............................................................ 14 2.2.1

  • Characterization of the Advanced Plant Experiment (APEX) passive residual heat removal system heat exchanger

    Creator:
    Stevens, Owen L.
    Abstract:
    The Oregon State University (OSU) Radiation Center (RC) is the location of a one quarter scale model of the Westinghouse Electric Corporation advanced light-water nuclear reactor design called the AP-600. The full scale AP-600 is a 600 megawatt electric nuclear power plant that incorporates unique passive systems to perform the...
    Resource Type:
    Masters Thesis
    Full Text:
    and Control 11 3. Hydraulic Characterization of the PRHR Hx. 13 3.1 Test Rig Description 14

  • Performance and stability analysis of gas-injection enhanced natural circulation in heavy-liquid-metal-cooled systems

    Creator:
    Yoo, Yeon-Jong
    Abstract:
    The purpose of this study is to investigate the performance and stability of the gas-injection enhanced natural circulation in heavy-liquid-metal-cooled systems. The target system is STAR-LM, which is a 400-MWt-class advanced lead-cooled fast reactor under development by Argonne National Laboratory and Oregon State University. The primary loop of STAR-LM relies...
    Resource Type:
    Dissertation

  • Feasibility of neutron radiographing reactor fuel assemblies

    Creator:
    Oien, C. T. (Charles T.)
    Abstract:
    Verification of fast reactor safety requires experiments wherein fuel bundles, enclosed in massive steel capsules, are subjected to simulated accident conditions in reactors such as TREAT(16) at ANL Idaho. To date, only small bundles (one to seven pins) have been tested in the experimental series. Determination of the fuel configuration...
    Resource Type:
    Masters Thesis
    Full Text:
    III. Neutron Source 13 Introduction 13 Reactor and Beam Layout 14 Beam Characteristics 18 Beam

  • Flat Plate Free Convective Heat Transfer

    Creator:
    Bristol, Benjanmin R.
    Abstract:
    The OSU MASLWR Test facility is a scaled integral systems LOCA and proof-of-concept test facility for the Multi-Application Small Light Water Reactor design concept. A unique feature of this design is to utilize a small high pressure containment vessel to provide the safety containment function for an integral reactor pressure...
    Resource Type:
    Masters Thesis
    Full Text:
    ] . . . . . . . . . . . . . . . . . . . . . 12 1.7 NIST Facility Piping Diagram. . . . . . . . . . . . . . . . . . . . . . . . . . . 14 1.8

  • Monte Carlo uncertainty reliability and isotope production calculations for a fast reactor

    Creator:
    Miles, Todd L.
    Abstract:
    With the advent of more powerful, less expensive computing resources, more and more attention is being given to Monte Carlo techniques in design application. In many circles, stochastic solutions are considered the next best thing to experimental data. Statistical uncertainties in Monte Carlo calculations are typically determined by the first...
    Resource Type:
    Masters Thesis
    Full Text:
    of neutron histories and the volume or 14 area of the cell. The SD* card allows the user to set

  • An assessment of thermal hydraulic analysis methods for pressurized thermal shock evaluations

    Creator:
    Young, Eric P.
    Abstract:
    Improved methods of determining temperature transients in reactor systems are desired because of recent interest in Pressurized Thermal Shock (PTS) issues. The research presented herein was performed in support of the Nuclear Regulatory Commission's effort to re-evaluate its existing PTS rules. These rules are particularly important to the re-licensing of...
    Resource Type:
    Masters Thesis
    Full Text:
    ................ 13 2-2 PLUME CENTERLINE TEMPERATURE........................................... 14 2-3 CREARE 1/5

  • Heat conduction in eccentric annuli

    Creator:
    Ting, Yine-ping
    Abstract:
    Muitiregion annular geometry is a basic feature of energy removal systems ranging from advanced heat exchangers to fission reactor fuel rods. For the symmetric cases, the heat conduction through annular regions is well defined. However, for a shift of an annular region, analysis of the system must be made in...
    Resource Type:
    Dissertation
    Full Text:
    convective boundary condition on the outer tube surface 14. Sensitivity study of the code-EXACT on the outer

  • On Primary Vortex Shedding Regimes between Long Tandem Plates

    Creator:
    Howard, Trevor Kent
    Abstract:
    Vortex shedding is a phenomenon relevant to any industry dealing with fluid flow. The shed vortices often produce oscillatory forces, which have been suspect in the catastrophic failure of airplanes and bridges alike. To prevent further engineering failures a better understanding of the underlying physics is needed. It has been...
    Resource Type:
    Dissertation
    Full Text:
    . ...................................... 14 Figure 2.2: Igarashi’s flow map of tandem cylinders

  • On the effects of centrifugal forces in air-water two-phase flow regime transitions of an adiabatic helical geometry

    Creator:
    Young, Eric P.
    Abstract:
    Two-phase flow in helical conduits is important in many industries where reaction between components, heat transfer, and mass transport are utilized as processes. The helical design is chosen for the effects of secondary flow patterns that reduce axial dispersion, increased heat transfer, and also their compact design. The first is...
    Resource Type:
    Dissertation
    Full Text:
    important tool of the averaging used in two-phase flow, short, and 14 interesting, so it is quickly

  • Neutron transport benchmarks for binary stochastic multiplying media : planar geometry, two energy groups

    Creator:
    Davis, Ian M. (Ian Mack)
    Abstract:
    Benchmark calculations are performed for neutron transport in a two material (binary) stochastic multiplying medium. Spatial, angular, and energy dependence are included. The problem considered is based on a fuel assembly of a common pressurized water nuclear reactor. The mean chord length through the assembly is determined and used as...
    Resource Type:
    Masters Thesis
    Full Text:
    ............................ 75 13 k-Eigenvalue PDF Calculation 1.1.1 ................. 79 14 k-Eigenvalue PDF Calculation 1.1.2

  • Stability improvement of the one-dimensional two-fluid model for horizontal two-phase flow with model unification

    Creator:
    Abel, Kent C.
    Abstract:
    The next generation of nuclear safety analysis computer codes will require detailed modeling of two-phase fluid flow. The most complete and fundamental model used for these calculations is known as the two-fluid model. It is the most accurate of the two-phase models since it considers each phase independently and links...
    Resource Type:
    Dissertation
    Full Text:
    Interfacial area (2.12) L Mixture volume 14 where L is the length scale at the interface and a1 is the

  • A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication

    Creator:
    Azarfar, Mohammad Ali
    Abstract:
    This study focuses on the development of a methodology to evaluate safeguardability of various processes in the nuclear fuel cycle. National and international implementation of safeguards systems has been investigated. It was concluded that proliferation is purely a political question that, hopefully, can be prevented by international norms and diplomacy;...
    Resource Type:
    Dissertation
    Full Text:
    Primary Safeguard Characters (case 6) 226 7-14 Process-Dependent Safeguard Characters and Their Relative

  • Numerical techniques for coupled neutronic/thermal hydraulic nuclear reactor calculations

    Creator:
    Betts, Curt M.
    Abstract:
    The solution of coupled neutronic/thermal hydraulic nuclear reactor calculations requires the treatment of the nonlinear feedback induced by the thermal hydraulic dependence of the neutron cross sections. As a result of these nonlinearities, current solution techniques often diverge during the iteration process. These instabilities arise due to the low level...
    Resource Type:
    Masters Thesis
    Full Text:
    . MODELING AND ASSUMPTIONS 14 A. General Treatment 14 B. Internal Heat Generation Calculation 18 C. Axial