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A New Innovative Spherical Cermet Nuclear Fuel Element to Achieve an Ultra-Long Core Life for use in Grid-Appropriate LWRs

Description: Spherical cermet fuel elements are proposed for use in the Atoms For Peace Reactor (AFPR-100) concept. AFPR-100 is a small-scale, inherently safe, proliferation-resistant reactor that would be ideal for deployment to nations with emerging economies that decide to select nuclear power for the generation of carbon-free electricity. The basic concept of the AFPR core is a water-cooled fixed particle bed, randomly packed with spherical fuel elements. The flow of coolant within the particle bed is a… more
Date: December 1, 2007
Creator: Senor, David J.; Painter, Chad L.; Geelhood, Ken J.; Wootan, David W.; Meriwether, George H.; Cuta, Judith M. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Assessment of Next Generation Nuclear Plant Intermediate Heat Exchanger Design.

Description: The Next Generation Nuclear Plant (NGNP), which is an advanced high temperature gas reactor (HTGR) concept with emphasis on production of both electricity and hydrogen, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 900-1000 C. In the indirect cycle system, an intermediate heat exchanger is used to transfer the heat from primary helium from the core to the secondary fluid, which can be helium, nitrogen/heliu… more
Date: October 17, 2008
Creator: Majumdar, S.; Moisseytsev, A. & Natesan, K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Experimental Validation of Stratified Flow Phenomena, Graphite Oxidation, and Mitigation Strategies of Air Ingress Accidents

Description: The US Department of Energy is performing research and development (R&D) that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP) Program / GEN-IV Very High Temperature Reactor (VHTR). Phenomena identification and ranking studies (PIRT) to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Schultz et al., 2006). Consequently, the development of advanced … more
Date: December 1, 2008
Creator: Oh, Chang Ho; Kim, Eung Soo; No, Hee Cheon & Cho, Nam Zin
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Argonne Liquid-Metal Advanced Burner Reactor : Components and In-Vessel System Thermal-Hydraulic Research and Testing Experience - Pathway Forward.

Description: This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor … more
Date: June 30, 2007
Creator: Kasza, K.; Grandy, C.; Chang, Y. & Khalil, H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Options to Extend the Applicability of High Temperature Irradiation Resistant Thermocouples

Description: Several options have been identified that could further enhance the reliability and increase the applicability of recently developed Idaho National Laboratory (INL) High Temperature Irradiation Resistant thermocouples (HTIR-TCs) for in-pile testing, allowing their use in higher temperature applications (up to at least 1700 °C). INL and the University of Idaho (UI) are investigating these options with the ultimate objective of providing recommendations for alternate thermocouple designs that are… more
Date: September 1, 2007
Creator: Rempe, Joy L.; Knudson, Darrell L.; Condie, Keith G.; Wilkins, S. Curtis; Crepeau, John C.; Daw, Joshua E. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

Description: This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant… more
Date: September 1, 2008
Creator: Mousseau, Vincent A.; Zhang, Hongbin & Zhao, Haihua
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

Description: The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation hea… more
Date: July 8, 2008
Creator: Revankar, S.T.; Zhou, W. & Henderson, Gavin
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Measurement of Turbulent Flow Phenomena for the Lower Plenum of a Prismatic Gas-Cooled Reactor

Description: Mean velocity field and turbulence data are presented for flow phenomena in a lower plenum of a typical prismatic gas-cooled reactor (GCR), such as in a Very High Temperature Reactor (VHTR) concept. In preparation for design, safety analyses and licensing, research has begun on readying the computational tools that will be needed to predict the thermal-hydraulics behavior of the reactor design. Fluid dynamics experiments have been designed and built to develop benchmark databases for the assess… more
Date: September 1, 2007
Creator: Jr., Hugh M. McIlroy; McEligot, Donald M.; Pink, Robert J.; Condie, Keith G. & McCreery, Glenn E.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

Description: Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially wher… more
Date: September 1, 2007
Creator: McCreery, Glenn & McIlroy, Hugh
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Status report on SHARP coupling framework.

Description: This report presents the software engineering effort under way at ANL towards a comprehensive integrated computational framework (SHARP) for high fidelity simulations of sodium cooled fast reactors. The primary objective of this framework is to provide accurate and flexible analysis tools to nuclear reactor designers by simulating multiphysics phenomena happening in complex reactor geometries. Ideally, the coupling among different physics modules (such as neutronics, thermal-hydraulics, and str… more
Date: May 30, 2008
Creator: Caceres, A.; Tautges, T. J.; Lottes, J.; Fischer, P.; Rabiti, C.; Smith, M. A. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Interim Report on Fuel Cycle Neutronics Code Development.

Description: As part of the Global Nuclear Energy Partnership (GNEP), a fast reactor simulation program was launched in April 2007 to develop a suite of modern simulation tools specifically for the analysis and design of sodium cooled fast reactors. The general goal of the new suite of codes is to reduce the uncertainties and biases in the various areas of reactor design activities by enhanced prediction capabilities. Under this fast reactor simulation program, a high-fidelity deterministic neutron transpor… more
Date: May 13, 2008
Creator: Rabiti, C; Smith, M. A.; Kaushik, D. & Yang, W. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Radioactive Isotope Production for Medical Applications Using Kharkov Electron Driven Subcritical Assembly Facility.

Description: Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an accelerator driven subcritical assembly. The main functions of the subcritical assembly are the medical isotope production, neutron thereby, and the support of the Ukraine nuclear industry. Reactor physics experiments and material research will be carried out using the capabilities of this facility. The United States of America and Ukraine have started collaboration activity for developing a conceptual desi… more
Date: May 15, 2007
Creator: Talamo, A. & Gohar, Y.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

An Innovative Hybrid Loop-Pool SFR Design and Safety Analysis Methods: Today and Tomorrow

Description: Investment in commercial sodium cooled fast reactor (SFR) power plants will become possible only if SFRs achieve economic competitiveness as compared to light water reactors and other Generation IV reactors. Toward that end, we have launched efforts to improve the economics and safety of SFRs from the thermal design and safety analyses perspectives at Idaho National Laboratory. From the thermal design perspective, an innovative hybrid loop-pool SFR design has been proposed. This design takes ad… more
Date: April 1, 2008
Creator: Zhang, Hongbin; Zhao, Haihua & Mousseau, Vincent
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Streamlining of the RELAP5-3D Code

Description: RELAP5-3D is widely used by the nuclear community to simulate general thermal hydraulic systems and has proven to be so versatile that the spectrum of transient two-phase problems that can be analyzed has increased substantially over time. To accommodate the many new types of problems that are analyzed by RELAP5-3D, both the physics and numerical methods of the code have been continuously improved. In the area of computational methods and mathematical techniques, many upgrades and improvements … more
Date: November 1, 2007
Creator: Mesina, George L; Hykes, Joshua & Guillen, Donna Post
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

Description: The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent ope… more
Date: June 1, 2008
Creator: Chang, G. S.; Lillo, M. A. & Ambrosek, R. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model

Description: The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that w… more
Date: September 1, 2007
Creator: Mousseau, Vincent A. & Pope, Michael A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

Description: The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power … more
Date: April 1, 2008
Creator: Oh, Chang H.; Kim, Eung Soo & Sherman, Steven
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Neutronics, Steady-State, and Transient Analyses for the Poland Maria Reactor for Irradiation Testing of Leu Lead Test Fuel Assemblies From Cerca : Anl Independent Verification Results.

Description: The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least… more
Date: June 7, 2011
Creator: Garner, P. L. & Hanan, N. A. (Nuclear Engineering Division)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Advanced High Temperature Reactor Systems and Economic Analysis

Description: The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support eit… more
Date: September 1, 2011
Creator: Holcomb, David Eugene; Peretz, Fred J & Qualls, A L
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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PRISMATIC CORE COUPLED TRANSIENT BENCHMARK

Description: The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivate… more
Date: June 1, 2011
Creator: Ortensi, J.; Pope, M.A.; Strydom, G.; Sen, R.S.; DeHart, M.D.; Gougar, H.D. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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AFIP-6 Breach Assessment Report

Description: Analysis of the AFIP-6 experiment is summarized in this report in order to determine the cause of gaseous fission product release observed during irradiation. During the irradiation, a series of small fission product releases were observed. In order to limit the potential for primary coolant contamination, the operating cycle was terminated and the AFIP-6 experiment was removed for examination. Both in-canal and post-irradiation examination revealed the presence of an unusually thick oxide laye… more
Date: February 1, 2011
Creator: Wachs, Dan; Robinson, Adam & Medvedev, Pavel
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

Description: The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used… more
Date: April 4, 2012
Creator: Wilson, E. H.; Horelik, N. E.; Dunn, F. E.; Newton, T. H., Jr.; Hu, L.; Stevens, J. G. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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DYNAMIC MODELING STRATEGY FOR FLOW REGIME TRANSITION IN GAS-LIQUID TWO-PHASE FLOWS

Description: In modeling gas-liquid two-phase flows, the concept of flow regime has been used to characterize the global interfacial structure of the flows. Nearly all constitutive relations that provide closures to the interfacial transfers in two-phase flow models, such as the two-fluid model, are often flow regime dependent. Currently, the determination of the flow regimes is primarily based on flow regime maps or transition criteria, which are developed for steady-state, fully-developed flows and widely… more
Date: September 1, 2011
Creator: Wang, X.; Sun, X. & Zhao, H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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MODELING STRATEGIES TO COMPUTE NATURAL CIRCULATION USING CFD IN A VHTR AFTER A LOFA

Description: A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) of the U.S. Department of Energy, Office of Nuclear Energy. In the design of the prismatic VHTR, hexagonal shaped graphite blocks are drilled to allow insertion of fuel pins, made of compacted TRISO fuel particles, and coolant channels for the helium coolant. One of the concerns for the reactor design is the effects of a loss of flow accident (LOFA) where the coo… more
Date: November 1, 2012
Creator: Tung, Yu-Hsin; Johnson, Richard W.; Chieng, Ching-Chang & Ferng, Yuh-Ming
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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