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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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Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.

Description: The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the … more
Date: October 1, 2007
Creator: Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J. (Idaho National Laboratory, Idaho Falls, ID); Parma, Edward J., Jr.; Peters, Curtis D. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Recovery Discontinuous Galerkin Jacobian-free Newton-Krylov Method for all-speed flows

Description: There is an increasing interest to develop the next generation simulation tools for the advanced nuclear energy systems. These tools will utilize the state-of-art numerical algorithms and computer science technology in order to maximize the predictive capability, support advanced reactor designs, reduce uncertainty and increase safety margins. In analyzing nuclear energy systems, we are interested in compressible low-Mach number, high heat flux flows with a wide range of Re, Ra, and Pr numbers.… more
Date: July 1, 2008
Creator: Park, HyeongKae; Nourgaliev, Robert; Mousseau, Vincent & Knoll, Dana
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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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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Benchmark Evaluation of Plutonium Hemispheres Reflected by Steel and Oil

Description: During the period from June 1967 through September 1969 a series of critical experiments was performed at the Rocky Flats Critical Mass Laboratory with spherical and hemispherical plutonium assemblies as nested hemishells as part of a Nuclear Safety Facility Experimental Program to evaluate operational safety margins for the Rocky Flats Plant. These assemblies were both bare and fully or partially oil-reflected. Many of these experiments were subcritical with an extrapolation to critical config… more
Date: June 1, 2008
Creator: Bess, John Darrell
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Progress reports for Gen IV sodium fast reactor activities FY 2007.

Description: An important goal of the US DOE Sodium Fast Reactor (SFR) program is to develop the technology necessary to increase safety margins in future fast reactor systems. Although no decision has been made yet about who will build the next demonstration fast reactor, it seems likely that the construction team will include a combination of international companies, and the safety design philosophy for the reactor will reflect a consensus of the participating countries. A significant amount of experience… more
Date: October 4, 2007
Creator: Cahalan, J. E. & Tentner, A. M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Addendum to NuMI shielding assessment

Description: The original safety assessment and the Safety Envelope for the NuMI beam line corresponds to 400 kW of beam power. The Main Injector is currently capable of and approved for producing 500 kW of beam power2. However, operation of the NuMI beam line at 400 kW of power brings up the possibility of an occasional excursion above 400 kW due to better than usual tuning in one of the machines upstream of the NuMI beam line. An excursion above the DOE approved Safety Envelope will constitute a safety vi… more
Date: October 1, 2007
Creator: Vaziri, Kamran
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Specification of advanced safety modeling requirements (Rev. 0).

Description: The U.S. Department of Energy's Global Nuclear Energy Partnership has lead to renewed interest in liquid-metal-cooled fast reactors for the purpose of closing the nuclear fuel cycle and making more efficient use of future repository capacity. However, the U.S. has not designed or constructed a fast reactor in nearly 30 years. Accurate, high-fidelity, whole-plant dynamics safety simulations will play a crucial role by providing confidence that component and system designs will satisfy establishe… more
Date: June 30, 2008
Creator: Fanning, T. H. & Tautges, T. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Final Report-Passive Safety Optimization in Liquid Sodium-Cooled Reactors.

Description: This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductio… more
Date: August 13, 2007
Creator: Cahalana, J. E. & Hahn, D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

A Review of Information for Managing Aging in Nuclear Power Plants

Description: Age related degradation effects in safety related systems of nuclear power plants should be managed to prevent safety margins from eroding below the acceptable limits provided in plant design bases. The Nuclear Plant Aging Research (NPAR) Pro- gram, conducted under the auspices of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, and other related aging management programs are developing technical information on managing aging. The aging management process cen… more
Date: September 1, 1995
Creator: Morgan, WC & Livingston, JV
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Methods for Quantifying Uncertainty in Fast Reactor Analyses.

Description: Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant … more
Date: April 7, 2008
Creator: Fanning, T. H. & Fischer, P. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN C-REACTOR DISASSEMBLY BASIN

Description: C-reactor disassembly basin is being prepared for deactivation and decommissioning (D and D). D and D activities will consist primarily of immobilizing contaminated scrap components and structures in a grout-like formulation. The disassembly basin will be the first area of the C-reactor building that will be immobilized. The scrap components contain aluminum alloy materials. Any aluminum will corrode very rapidly when it comes in contact with the very alkaline grout (pH > 13), and as a resul… more
Date: July 12, 2011
Creator: Wiersma, B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Ion Desorption Stability in Superconducting High Energy Physics Proton Colliders

Description: In this paper we extend our previous analysis of cold beam tube vacuum in a superconducting proton collider to include ion desorption in addition to thermal desorption and synchrotron radiation induced photodesorption. The new ion desorption terms introduce the possibility of vacuum instability. This is similar to the classical room temperature case but now modified by the inclusion of ion desorption coefficients for cryosorbed (physisorbed) molecules which can greatly exceed the coefficients f… more
Date: May 29, 1995
Creator: Turner, W.C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Heartbeat Model for Component Failure in Simulation of Plant Behavior

Description: As part of the Department of Energy’s “Light Water Reactor Sustainability Program” (LWRSP), tools and methodology for risk-informed characterization of safety margin are being developed for use in supporting decision-making on plant life extension after the first license renewal. Beginning with the traditional discussion of “margin” in terms of a “load” (a physical challenge to system or component function) and a “capacity” (the capability of that system or component to accommodate the challeng… more
Date: March 1, 2011
Creator: Youngblood, R. W.; Nourgaliev, R. R.; Kelly, D. L.; Smith, C. L. & Dinh, T-N.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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INTEGRATION OF RELIABILITY WITH MECHANISTIC THERMALHYDRAULICS: REPORT ON APPROACH AND TEST PROBLEM RESULTS

Description: The Risk-Informed Safety Margin Characterization (RISMC) pathway of the Light Water Reactor Sustainability Program is developing simulation-based methods and tools for analyzing safety margin from a modern perspective. [1] There are multiple definitions of 'margin.' One class of definitions defines margin in terms of the distance between a point estimate of a given performance parameter (such as peak clad temperature), and a point-value acceptance criterion defined for that parameter (such as 2… more
Date: July 1, 2011
Creator: Schroeder, J. S. & Youngblood, R. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

Description: The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the curren… more
Date: May 1, 2012
Creator: Andrs, David; Berry, Ray; Gaston, Derek; Martineau, Richard; Peterson, John; Zhang, Hongbin et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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PROPERTIES IMPORTANT TO MIXING FOR WTP LARGE SCALE INTEGRATED TESTING

Description: Large Scale Integrated Testing (LSIT) is being planned by Bechtel National, Inc. to address uncertainties in the full scale mixing performance of the Hanford Waste Treatment and Immobilization Plant (WTP). Testing will use simulated waste rather than actual Hanford waste. Therefore, the use of suitable simulants is critical to achieving the goals of the test program. External review boards have raised questions regarding the overall representativeness of simulants used in previous mixing tests.… more
Date: April 26, 2012
Creator: Koopman, D.; Martino, C. & Poirier, M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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SHIPMENT OF NON-TRADITIONAL CONTENTS IN THE 9977 TYPE B PACKAGE

Description: The 9977 is a certified Type B Packaging authorized to ship uranium and plutonium in metal and oxide forms. These materials are typically confined within metallic containers designed for ease of handling and to prevent the spread of contamination. The Pacific Northwest National Laboratory (PNNL) uses Pu and U sources for the training of domestic and international customs agents in the identification and detection of radioactive materials (RAM). These materials are packed in polycarbonate contai… more
Date: June 6, 2011
Creator: Abramczyk, G.; Loftin, B.; Bellamy, S. & Nathan, S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry.

Description: To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and… more
Date: May 23, 2011
Creator: Tzanos, C. P. & Dionne, B. (Nuclear Engineering Division)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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ESTIMATED SIL LEVELS AND RISK COMPARISONS FOR RELIEF VALVES AS A FUNCTION OF TIME-IN-SERVICE

Description: Risk-based inspection methods enable estimation of the probability of spring-operated relief valves failing on demand at the United States Department of Energy's Savannah River Site (SRS) in Aiken, South Carolina. The paper illustrates an approach based on application of the Frechet and Weibull distributions to SRS and Center for Chemical Process Safety (CCPS) Process Equipment Reliability Database (PERD) proof test results. The methodology enables the estimation of ANSI/ISA-84.00.01 Safety Int… more
Date: March 26, 2012
Creator: Harris, S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

Description: This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradat… more
Date: March 1, 2012
Creator: Renfro, David G; Cook, David Howard; Freels, James D; Griffin, Frederick P; Ilas, Germina; Sease, John D et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

Description: The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D&D). D&D activities will consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS&am… more
Date: October 29, 2009
Creator: Wiersma, B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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NEAMS Experimental Support for Code Validation, INL FY2009

Description: 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 where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Whereas the Verification part of the process does not rely on experiment, the Validation part, on the contrary, necessitates as many relevant and pre… more
Date: September 1, 2009
Creator: Youinou, G.; Palmiotti, G.; Salvatore, M. & Rabiti, C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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