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Sample records for water reactors casl

  1. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Media Kit CASL Acknowledgement This research was supported by the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy Contract No. DE-AC05-00OR22725. CASL Logo Files CASL Extended - CASL_word.jpg and CASL_word.png CASL without words - CASL.jpg and CASL.png CASL with words - CASL_word.jpg and CASL_word.png CASL Partners - partners.jpg

  2. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    J.C., CASL: Consortium for the Advanced Simulation of Light Water Reactors - A DOE Energy Innovation Hub, ANS MC2015 Joint Internation Conference on Mathematics and Computation...

  3. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Science and Technology Archive Energy Department Announces Five Year Renewal of Funding for First Energy Innovation Hub Consortium for Advanced Simulation of Light Water Reactors to Receive up to $121.5 Million Over Five Years. Posted: January 29, 2015 VERA-CS Coupled Multi-physics Capability demonstrated in a Full Core Simulation In December, CASL reported on the latest results from its Watts Bar reactor progression problem modeling. Posted: August 14, 2014 Westinghouse Completes its AP1000®

  4. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    failure in a commercial power reactor, CASL Technical Report: ... and Simulation of the AP1000 PWR Cycle 1 Depletion, ... Framework for Design Optimization, Parameter ...

  5. CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    AMA.NRC.P5.01 CASL NRC Commissioner Technical Seminar Jess Gehin Oak Ridge National Laboratory December 22, 2012 CASL-U-2014-0076-000-a CASL-U-2012-0076-000-a 1 CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub for Modeling and Simulation of Nuclear Reactors NRC Commissioner Technical Seminar November 30, 2012 Doug Kothe (ORNL) CASL Director Doug Burns (INL) CASL Deputy Director Paul Turinsky (NCSU) CASL Chief Scientist Jess Gehin (ORNL) CASL AMA FA

  6. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes … September 9, 2010

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    1 CASL Industry Council Meeting March 7 - 8, 2012 - Raleigh, NC Minutes The fourth meeting of the Industry Council (IC) for the Consortium for Advanced Simulation of Light Water Reactors (CASL) was held on March 7 until noon March 8, 2012, at the CASL Facility of North Carolina State University, Raleigh, NC. The meeting was chaired by John Gaertner of EPRI. Attendance was by invitation only. Fifteen representatives from 14 of the 19 member organizations attended. One guest from Nuclear Energy

  7. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    on issues related to management, performance, strategic direction, and institutional interfaces within CASL. The CASL Director reports to the BOD on all matters related to CASL...

  8. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Back Board of Directors The CASL Board of Directors (BOD) serves as both an advisory and oversight body for the ORNL Laboratory Director and the CASL Senior Leadership Team (SLT) on issues related to management, performance, strategic direction, and institutional interfaces within CASL. The CASL Director reports to the BOD on all matters related to CASL strategic program plans and decisions. The BOD works to ensure the execution of CASL operational and R&D plans provide maximum benefit to

  9. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    in Drekar, CASL Technical Report: CASL-U-2012-0080-000, June 30, 2012. Bakosi, J., N. Barnett, M.A. Christon, M.M. Francois, R.B. Lowrie and R. Sankaran, Integration of Hydra-TH...

  10. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Back Science Council The Science Council The Science Council provides independent assessment of whether the CASL scientific work planned and executed is of high quality and supports attaining the goals of CASL. In addition, the Science Council may be called upon to complete detailed assessments of specific CASL scientific issues. The Science Council advises the following CASL Focus Areas (FAs): Radiation Transport Methods (RTM), Thermal Hydraulics Methods (THM), Materials Performance and

  11. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Organization The CASL organizational structure (see chart) has proven to accommodate necessary program priority changes and risk management actions during CASL's lifetime, yet possesses a primary structure that is stable and functional. Major features include: Central, integrated management working predominately from a single location at ORNL: Director with full line authority and accountability for all aspects of CASL; Deputy Direct to drive program planning, performance and assessment; Chief

  12. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Site Map Home CASL Partners Research Science & Technology Archive Journal & Conference Papers Technical Reports Presentations VERA Software & Support VERA 3.3 VERA.edu How To...

  13. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    of technology. Management Performance reflects CASL's ability to meet its virtual one-roof plan (collocation), maintain consortium cohesion and chemistry, and deliver its...

  14. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    modeling and simulation technology that is deployed and applied broadly throughout the nuclear energy industry to enhance safety, reliability, and economics. CASL will address,...

  15. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    plant power uprates, life extension, and higher burnup fuels Provide the primary bridge between the scientific and computational capabilities developed by CASL and external...

  16. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL's Latest Research CIPS Simulation Capability Implemented in VERA Posted on October 28, 2015 Departure from Nucleate Boiling (DNB) Multi-Physics Approach & Applications using...

  17. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    from CASL (Dr. Mike Short, MIT, October, 31, 2013) Subchannel Methods for the Thermal-Hydraulic Analysis for Nuclear Power Systems (Dr. Michael Doster, NCSU, May 28, 2013)...

  18. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Site Map Home About CASL Vision Mission Goals Strategy Integration Performance Metrics Partners Founding Partners Electric Power Research Institute Idaho National Laboratory Los...

  19. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes … September 9, 2010

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    September 11 - 12, 2012 - Oak Ridge, TN Minutes The fifth meeting of the Industry Council (IC) for the Consortium for Advanced Simulation of Light Water Reactors (CASL) was held on September 11 and 12, 2012; at Oak Ridge National Laboratory in Oak Ridge, TN. The first day was a joint meeting of the CASL Science Council and the Industry Council and was co-facilitated by Paul Turinsky of NCSU and John Gaertner of EPRI. The Industry Council met separately on the second day which was chaired by John

  20. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Engineering and Design, Volume Online, Issue CASL-U-2015-0301-000, August 28, 2015. Smith, T.M., M.A. Christon, E. Baglietto and H. Luo, "Assessment of Models for Near Wall...

  1. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Upcoming Training Events CASL Events Your browser does not appear to support JavaScript, but this page needs to use JavaScript to display correctly. You can visit the HTML-only...

  2. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    capabilities to meet future CASL needs. DTK has been given an open source BSD 3-clause license. The primary code development repository is publicly-hosted under the GitHub group...

  3. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL Partners Electric Power Research Institute Idaho National Laboratory Los Alamos National Laboratory Massachusetts Institute of Technology North Carolina State University Oak Ridge National Laboratory Sandia National Laboratory Tennessee Valley Authority University of Michigan Westinghouse Electric Company

  4. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Back Industry Council Chairperson: Scott Thomas, Duke Energy Executive Director: Erik Mader, EPRI Mission and Objectives The mission of the Industry Council (IC) is to ensure that CASL solutions are "used and useful", and that CASL provides effective leadership advancing the Modeling and Simulation state-of-the art in the nuclear industry. Specific objectives of the Industry Council are: Early, continuous, and frequent interface and engagement of end-users and technology providers

  5. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Video Gallery Management of Uncertainties in Predictive Science presented by Dr. Hany Abdel-Khalik and Dr. Ralph Smith, NCSU. Surrogate Models for Uncertainty Quantification presented by Dr. Ralph Smith, NCSU. Subchannel Methods for the Thermal-Hydraulic Analysis for Nuclear Power Systems presented by Dr. Michael Doster, NCSU. Finding the Cure for CRUD: Insights from CASL presented by Dr. Mike Short, MIT. Andrew Godfrey (ORNL) describes CASL -- the Consortium for Advanced Simulation of Light

  6. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    It informs consequential nuclear power operational and ... Opportunities for reduced uncertainties in design and ... to address light water reactor (LWR) operational and ...

  7. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    ... (IAEA) Meeting on Accident Tolerant Fuel (ATF) Concepts for LWRs, International Atomic ... Simulation of Light Water Reactors, Westinghouse Technology Users Group, August 28, ...

  8. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Virtual Environment for Scientific Collaboration Posted: April 30, 2013 The Consortium for Advanced Simulation of Light Water Reactors, the Department of Energy's first...

  9. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes … September 9, 2010

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Meetings January 11, 2011 - Oak Ridge, TN February 8, 2011 - Charlotte, NC Minutes The second meeting of the Industry Council (IC) for the Consortium for Advanced Simulation of Light Water Reactors (CASL) was held in two parts on January 11, 2011 at Oak Ridge National Laboratories (ORNL), Oak Ridge, TN; and on February 8, 2011, at the facilities of the Electric Power Research Institute (EPRI) in Charlotte, NC. Both meetings were chaired by John Gaertner of EPRI. Two meetings were necessary

  10. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes … September 9, 2010

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    August 23 - 24, 2011 - Oak Ridge, TN Minutes The third meeting of the Industry Council (IC) for the Consortium for Advanced Simulation of Light Water Reactors (CASL) was held on August 23 until noon on August 24, 2011, at Oak Ridge National Laboratories (ORNL), Oak Ridge, Tennessee. The meeting was chaired by John Gaertner of EPRI. The agenda, meeting attendees, and IC member organizations are included in Attachment 1 to these minutes. Attendance was by invitation only. Fifteen representatives

  11. CONSORTIUM FOR ADVANCED SIMULATION OF LIGHT WATER REACTORS (CASL) Meeting Notes … September 9, 2010

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Meeting September 9, 2010 Minutes The first meeting of the Industry Council (IC) for the Consortium for Advanced Simulation of Light Water Reactors (CASL) was held on September 9, 2010, at the facilities of the Electric Power Research Institute (EPRI) in Charlotte, NC. The meeting was chaired by John Gaertner of EPRI. The meeting attendees and their affiliations are listed on Attachment 1 to these minutes. Attendance was by invitation only. Representatives from 16 organizations were invited. All

  12. CASL-U-2015-0248-000 Modeling Boiling Water Reactor

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    8-000 Modeling Boiling Water Reactor Designs using MPACT Andrew P. Fitzgerald Brendan Kochunas Daniel Jabaay Thomas Downar University of Michigan July 7, 2015 CASL-U-2015-0248-000 ATRIUM TM 10: K-inf vs burn-up for the ATRIUM TM 10 lattice from various transport codes. MPACT is shown to have the ability to model some BWR features such as (square) channel boxes, water rods, and water channels with reasonable accuracy. The ATRIUM TM 10 comparison has shown MPACT can predict k-inf with similar

  13. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    reactor physical phenomena using coupled multiphysics models. VERA also includes the software development environment and computational infrastructure needed for these...

  14. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    to achieve challenge problem solutions A strong VERA infrastructure supporting software development, testing, and releases. Requirements Drivers Modeling of reactors...

  15. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Radiation Transport Methods (RTM) Delivers next-generation radiation transport tools to the virtual Reactor RTM Vision Statement Objectives and Strategies Next generation,...

  16. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    LWRs; Develop and effectively apply modern virtual reactor technology; Engage the nuclear energy community through modeling and simulation; and Deploy new partnership and...

  17. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Thermal Hydraulics Methods (THM) Delivers next-generation thermal-hydraulic simulation tools to Virtual Environment for Reactor Applications (VERA) Thermal Hydraulics Methods...

  18. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Read More Fuel pellets and the fuel rod cladding are affected at the microstructural level Under pressure, heat and radiation in the reactor environment, both the fuel pellets and ...

  19. ORNL). Consortium for Advanced Simulation of Light Water Reactors

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Simulation of Light Water Reactors (CASL) was established by the US Department of Energy in 2010 to advance modeling and simulation capabilities for nuclear reactors. CASL's...

  20. COLLOQUIUM: CASL: Consortium for Advanced Simulation of Light Water

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Reactors, a DOE Energy Innovation Hub | Princeton Plasma Physics Lab May 29, 2013, 4:15pm to 5:30pm Colloquia MBG Auditorium COLLOQUIUM: CASL: Consortium for Advanced Simulation of Light Water Reactors, a DOE Energy Innovation Hub Dr. Douglas Kothe Oak Ridge National Laboratory The Consortium for Advanced Simulation of Light Water Reactors (CASL) is the first U.S. Department of Energy (DOE) Energy Innovation Hub, established in July 2010 for the modeling and simulation (M&S) of nuclear

  1. Some Specific CASL Requirements for Advanced Multiphase Flow Simulation of Light Water Reactors

    SciTech Connect

    R. A. Berry

    2010-11-01

    Because of the diversity of physical phenomena occuring in boiling, flashing, and bubble collapse, and of the length and time scales of LWR systems, it is imperative that the models have the following features: • Both vapor and liquid phases (and noncondensible phases, if present) must be treated as compressible. • Models must be mathematically and numerically well-posed. • The models methodology must be multi-scale. A fundamental derivation of the multiphase governing equation system, that should be used as a basis for advanced multiphase modeling in LWR coolant systems, is given in the Appendix using the ensemble averaging method. The remainder of this work focuses specifically on the compressible, well-posed, and multi-scale requirements of advanced simulation methods for these LWR coolant systems, because without these are the most fundamental aspects, without which widespread advancement cannot be claimed. Because of the expense of developing multiple special-purpose codes and the inherent inability to couple information from the multiple, separate length- and time-scales, efforts within CASL should be focused toward development of a multi-scale approaches to solve those multiphase flow problems relevant to LWR design and safety analysis. Efforts should be aimed at developing well-designed unified physical/mathematical and high-resolution numerical models for compressible, all-speed multiphase flows spanning: (1) Well-posed general mixture level (true multiphase) models for fast transient situations and safety analysis, (2) DNS (Direct Numerical Simulation)-like models to resolve interface level phenmena like flashing and boiling flows, and critical heat flux determination (necessarily including conjugate heat transfer), and (3) Multi-scale methods to resolve both (1) and (2) automatically, depending upon specified mesh resolution, and to couple different flow models (single-phase, multiphase with several velocities and pressures, multiphase with single

  2. CASL Photo Book

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL Photo Book August 7, 2014 CASL-U-2014-0177-000 CASL-U-2014-0177-000 Predictive science-based simulation technology that harnesses world-class computational power to enhance clean energy production CASL-U-2014-0177-000 Cover photo: 3D visualizations allow a physical walk-through of the top 20% of high-powered rods in a pressurized water reactor core. Details revealed through such visualizations can provide insight into factors affecting core performance. CASL team members often review and

  3. CASL-U-2015-0040-000 Initial Boiling Water

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    40-000 Initial Boiling Water Reactor (BWR) Input Specifications Scott Palmtag Core Physics February 28, 2015 Initial Boiling Water Reactor (BWR) Input Specification Consortium for Advanced Simulation of LWRs ii CASL-U-2015-0040-000 REVISION LOG Revision Date Affected Pages Revision Description 0 02/28/2015 All Original Report for L3:PHI.VCS.P10.02 Document pages that are: Export Controlled NO IP/Proprietary/NDA Controlled NO Sensitive Controlled NO Requested Distribution: To: Copy: Initial

  4. CASL - CASL and TVA

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL and TVA The Tennessee Valley Authority (TVA) provides operational experience and validation information to CASL. Many of the CASL staff are highly skilled in their nuclear energy-relevant area of expertise, but not necessarily operational experience typified by the day-to-day issues encountered in commercial reactors. For example, the physical process of removing the fuel from the core, inspecting it, and then repacking it for the next cycle, which for TVA is planned for months in advance

  5. CASL Industry Council Meeting

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Meeting 12-13 April 2016 Meeting Minutes Page | 1 The spring 2016 meeting of the Industry Council (IC) for the Consortium for Advanced Simulation of Light Water Reactors (CASL) was held on April 12-13, 2016 at the Aloft Hotel in Greenville, South Carolina and was led by the CASL IC Chairman Scott Thomas of Duke Energy and the new CASL IC Executive Director Erik Mader from the EPRI Fuel Reliability Program. The meeting location and logistics were excellent and the group profusely thanked Lorie

  6. CASL - Image Gallery

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Image Gallery All Works CASL Modeling Applications Multi-Physics Neutronics Thermal Hydraulics Fuel Performance Corrosion Chemistry Secretary Moniz tours the Consortium... Secretary Moniz tours the Consortium for Advanced Simulation of Light Water Reactors (CASL) View The all-quartz test section design allows... The all-quartz test section design allows for simultaneous measurement of the temperature and phase distribution on the boiling surface, as well as measurement of the velocity field in

  7. CASL-U-2015-0104-000 CASL: The Consortium for

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    4-000 CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub Doug Kothe Oak Ridge National Laboratory Paul Turinsky North Carolina State University July 8-10, 2013 CASL-U-2015-0104-000 1 CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub Doug Kothe and Paul Turinsky CASL-U-2015-0104-000 2 * A Different Approach - "Multi-disciplinary, highly collaborative teams ideally working under one roof to solve

  8. Thermal hydraulics development for CASL

    SciTech Connect

    Lowrie, Robert B

    2010-12-07

    This talk will describe the technical direction of the Thermal-Hydraulics (T-H) Project within the Consortium for Advanced Simulation of Light Water Reactors (CASL) Department of Energy Innovation Hub. CASL is focused on developing a 'virtual reactor', that will simulate the physical processes that occur within a light-water reactor. These simulations will address several challenge problems, defined by laboratory, university, and industrial partners that make up CASL. CASL's T-H efforts are encompassed in two sub-projects: (1) Computational Fluid Dynamics (CFD), (2) Interface Treatment Methods (ITM). The CFD subproject will develop non-proprietary, scalable, verified and validated macroscale CFD simulation tools. These tools typically require closures for their turbulence and boiling models, which will be provided by the ITM sub-project, via experiments and microscale (such as DNS) simulation results. The near-term milestones and longer term plans of these two sub-projects will be discussed.

  9. Los Alamos boosts light-water reactor research with advanced...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Los Alamos boosts light-water reactor research Los Alamos boosts light-water reactor research with advanced modeling and simulation technology As part of the consortium CASL will ...

  10. CASL - Industry Council and CASL Project are Engaged to Investigate

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Workflow Processes to Ensure Value and Usability of CASL Results Industry Council and CASL Project are Engaged to Investigate Workflow Processes to Ensure Value and Usability of CASL Results John Gaertner, Chairman, CASL Industry Council January 3, 2012 Background The CASL Workflow Project is a joint activity between members of the CASL Industry Council (IC) and the Advanced Modeling Applications (AMA) and Virtual Reactor Integration (VRI) focus areas. The project originated at a Fall 2011

  11. CASL Industry Council Meeting

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    6 CASL Industry Council Meeting March 26-27, 2013 - Cranberry Township, PA Minutes The sixth meeting of the Industry Council (IC) for the Consortium for Advanced Simulation of Light Water Reactors (CASL) was held on March 26-27, 2013 at Westinghouse in Cranberry Township, PA. The first day of the Industry Council was chaired by John Gaertner and the second day was chaired by Heather Feldman. The meeting attendees and their affiliations are listed on Attachment 1 to these minutes. Attendance was

  12. CASL Industry Council Meeting

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Industry Council Meeting 4 - 5 November 2015 Meeting Minutes The autumn 2015 meeting of the Industry Council (IC) for the Consortium for Advanced Simulation of Light Water Reactors (CASL) was held on 4 - 5 November 2015 at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, TN. The first day of meeting was a joint meeting of the CASL Industry and Science Councils and was held at the Spallation Neutron Source (SNS) facility at ORNL. An independent IC meeting was held the morning of the second

  13. Light Water Reactors A DOE Energy Innovation Hub for Modeling...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    and Simulation of Nuclear Reactors CASL is focused on three issues for nuclear energy: reducing cost, reducing the amount of used nuclear fuel, and safety. CASL core...

  14. CASL-U-2015-0278-000 CASL Program Highlights

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    78-000 CASL Program Highlights July 2015 Jess Gehin Oak Ridge National Laboratory July 31, 2015 CASL-U-2015-0278-000 1 Demonstrate on new VERA Boiling Water Reactor (BWR) Neutronics Capability * Objective of work is to implement and demonstrate modeling capability for BWR Control Cell * Stretch goals of capability for 2-D, Full Core BWR core and 3-D single assembly modeling demonstrated Plenum Natural U. Blanket Part Length Rod 2-D Core Demonstration 3-D Assembly Model Demonstration Fast Neutron

  15. CASL: Renewal Application Process

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Update Jess Gehin (ORNL), Director Doug Burns (INL), Deputy Director Dave Kropaczek (NCSU), Chief Scientist CASL Industry Council Meeting Greeneville, South Carolina April 13, 2016 CASL-U-2016-1081-000 2 CASL's Mission is to Provide Leading- Edge M&S Capabilities to Improve the Performance of Operating LWRs VISION Predict, with confidence, the performance and assured safety of nuclear reactors, through comprehensive, science-based M&S technology deployed and applied broadly by the U.S.

  16. Validation and Uncertainty Quantification in the Consortium for Advanced Simulation of Light Water Reactors

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    and Uncertainty Quantification in CASL Michael Pernice Center for Advanced Modeling and Simulation Idaho National Laboratory SAMSI Uncertainty Quantification Transition Workshop May 21-23 2012 CASL-U-2012-0108-000 What Is CASL? * Consortium for Advanced Simulation of LWRs - An Energy Innovation Hub * Objective: predictive simulation of light water reactors - Reduce capital and operating costs * Power uprates * Lifetime extension - Reduce nuclear waste * Higher fuel burnup - Enhance operational

  17. CASL - CASL's presence at Nureth-14

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL's presence at Nureth-14 CASL Thermal Hydraulics team had a considerable presence at the 14th International Topical Meeting on Nuclear Reactor Thermalhydraulics that took place in Toronto from Sept. 25-30, 2011. The NURETH series of conferences is an outstanding international technical forum on different topics related to thermalhydraulics and nuclear reactor safety with participation from leading practitioners and academic and industry researchers from around the world. NURETH-14 was

  18. CASL - Public Release of CASL Infrastructure Software

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Public Release of CASL Infrastructure Software The Virtual Environment for Reactor Applications (VERA) comprises a suite of tools for scalable simulation of nuclear reactor core behavior. The following three key components of the VERA infrastructure have been released and made publicly-available: Data TransferKit Data TransferKit (DTK) is a software package designed to provide grid transfer services to the various CASL physics components. The Data Transfer Kit has been redesigned to provide a

  19. CASL - Validation of Peregrine with Test Reactor Data

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Validation of Peregrine with Test Reactor Data At the end of September, Pellet-Cladding Interaction (PCI) Challenge Problem Integrator Robert Montgomery reported that good progress has been made in demonstrating the Peregrine LWR fuel performance modeling software. The Peregrine fuel performance analysis computer program is being developed to provide a single rod 3-dimensional fuel performance modeling capability to assess safety margins and the impact of plant operation and fuel rod design on

  20. CASL-U-2015-0073-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    73-000 The Consortium for Advanced Simulation of Light Water Reactors - A DOE Energy Innovation Hub Technical Society Meeting of Knoxville Jess C. Gehin Oak Ridge National Laboratory March 3, 2014 CASL-U-2015-0073-000 The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub Jess C. Gehin Oak Ridge National Laboratory Technical Society Meeting of Knoxville March 3, 2014 CASL-U-2015-0073-000 2 2 * A Different Approach - Multi-disciplinary, highly collaborative

  1. CASL - PWR Reactor Vessel Multi-Physics CFD Model

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    PWR Reactor Vessel Multi-Physics CFD Model Jin Yan*1, Yiban Xu1, Andrew Petrarca1, Zeses Karoutas1, Emre Tatli1, Emilio Baglietto2, Jess Gehin3 1Westinghouse Electric Company LLC 2Massachusetts Institute of Technology 3Oak Ridge National Lab *Correspondence to: yan3j@westinghouse.com A complete 3D SolidWorks CAD model of Watts Bar Unit 1 was constructed based on drawings. A single fuel assembly CAD model including all geometrical details was created based on the Westinghouse V5H 17x17 fuel

  2. User guidelines and best practices for CASL VUQ analysis using Dakota.

    SciTech Connect

    Adams, Brian M.; Swiler, Laura Painton; Hooper, Russell; Lewis, Allison; McMahan, Jerry A.,; Smith, Ralph C.; Williams, Brian J.

    2014-03-01

    Sandia's Dakota software (available at http://dakota.sandia.gov) supports science and engineering transformation through advanced exploration of simulations. Specifically it manages and analyzes ensembles of simulations to provide broader and deeper perspective for analysts and decision makers. This enables them to enhance understanding of risk, improve products, and assess simulation credibility. This manual offers Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) partners a guide to conducting Dakota-based VUQ studies for CASL problems. It motivates various classes of Dakota methods and includes examples of their use on representative application problems. On reading, a CASL analyst should understand why and how to apply Dakota to a simulation problem. This SAND report constitutes the product of CASL milestone L3:VUQ.V&V.P8.01 and is also being released as a CASL unlimited release report with number CASL-U-2014-0038-000.

  3. CASL - Breakthrough in Theory and Modeling of Dimensional Instability of

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Zr-Based Alloys in Light Water Reactors Breakthrough in Theory and Modeling of Dimensional Instability of Zr-Based Alloys in Light Water Reactors S.I. Golubov, R.E. Stoller - Oak Ridge National Laboroatory A.V. Barashev - University of Tennessee, Knoxville Generalization of radiation growth model developed at ORNL to describe deformation of fuel cladding materials under applied loads The Consortium for Advanced Simulation of Light Water Reactors (CASL) at ORNL is addressing a wide range of

  4. Consortium for Advanced Simulation of Light Water Reactors

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    An essential part of developing a closed form set of equations (closures) for prediction of two-phase flow with computational fluid dynamics (CFD) is understanding how the bubbles generat- ed by boiling interact. An accurate prediction of moderator and fuel performance once boiling has begun is needed to simulate CASL Challenge Problems related to boiling water reactors (BWRs), departure from nucleate boiling (DNB) behavior in pressurized water reactors (PWRs), loss of coolant accidents (LOCAs),

  5. CASL-8-2015-0047-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    47-000 Estimation of the Shear- Induced Lift Force on a Single Bubble in Laminar and Turbulent Shear Flows Using Interface Tracking Approach Aaron Matt Thomas North Carolina State University November 20, 2014 CASL-U-2015-0047-000 Estimation of the Shear-Induced Lift Force on a Single Bubble in Laminar and Turbulent Shear Flows Using Interface Tracking Approach Consortium for Advanced Simulation of Light Water Reactors Undergraduate Research presented by Aaron Matt Thomas Under the direction of

  6. CASL - Nuclear Energy - Supercomputer speeds path forward

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Nuclear Energy - Supercomputer speeds path forward When run on graphics processing units, Denovo ran 3.5 times faster than what was possible with ORNL's Jaguar, which uses only central processing units. With this increase, 3D simulations are now within reach, said Tom Evans, who led the Denovo development team. Titan is a GPU/CPU hybrid to be installed over the next several months. This research supports the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov/).

  7. CASL-U-2015-0249-000 Rod Cusping

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    9-000 Rod Cusping Treatment in MPACT Aaron Graham and Tom Downar University of Michigan Ben Collins Oak Ridge National Laboratory July 7, 2015 CASL-U-2015-0249-000 Rod Cusping Treatment in MPACT CASL is currently developing a new core simulator called MPACT to solve neutron transport problems for light- water nuclear reactors. MPACT uses the 2D/1D approach, which is an iterative method with two primary steps in each iteration. The first step consists of a series of high-fidelity calculation in

  8. CASL Milestone L2.VRI.P6.01

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    L2 Milestone VRI.P7.01 - VERA 3.1 Protected under CASL Master NDA Official Use Only CASL-U-2013-0164-000 i Consortium for Advanced Simulation of LWRs Fe Milestone L2:VRI.P7.01 Virtual Environment for Reactor Applications (VERA): Snapshot 3.1 07/24/2013 John A. Turner CASL-U-2013-0164-000 CASL L2 Milestone VRI.P7.01 - VERA 3.1 CASL-U-2013-0164-000 ii Consortium for Advanced Simulation of LWRs Oak Ridge National Laboratory in partnership with Electric Power Research Institute Idaho National

  9. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    codes (e.g,. a physics simulation) and iterative systems analysis methods such as optimization or uncertainty quantification. It includes algorithms for: optimization with...

  10. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    models are being developed based on higher fidelity CFD methods, and may also include adhesionstrength models16 for the crud's surface layer as well as other "release"...

  11. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    that have included management of CERCLA and RCRA remediation projects at the INL, Rocky Flats, and Mound laboratories, management of special nuclear materials at the INL, and...

  12. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Nuclear Energy Defense Waste Management Programs Advanced Nuclear Energy Nuclear Energy Safety Technologies Facilities Battery Abuse Testing Laboratory Cylindrical Boiling ...

  13. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    The toolkit also provides run-time parallel domain decomposition with data-migration for both static and dynamic load-balancing. Linear algebra is handled through an...

  14. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Office of Nuclear Energy (NE) for their advancement of nuclear power; U.S. Nuclear Regulatory Commission (NRC) for safety reviews and licensing; R&D community for identification,...

  15. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Contact Us Address Oak Ridge National Laboratory PO Box 2008, MS6003 Oak Ridge, TN 37831-6003 Email Information Support ORNL Campus...

  16. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    and Monte Carlo transport applications. Exnihilo is based on a package architecture model such that each package provides well-defined capabilities. Exnihilo currently...

  17. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    model and numerical algorithm requirements of VERA. THM collaborates closely with Materials Performance and Optimization (MPO) for sub-grid material and chemistry models,...

  18. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    integration (VRI) for integration and development of VERA. Materials Performance and Optimization (MPO) - Develops improved materials performance models for fuels, cladding,...

  19. Consortium for Advanced Simulation of Light Water Reactors (CASL)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    VERA VERA.edu Software availability U.S. Citizens and most LPRs as limited by U.S. export control regulations Students, Faculty MOC radiation transport included included Sn and SPn radiation transport included Not included Monte Carlo radiation transport included included Integrated cross-section library included Limited functionality Integrated depletion library included Limited functionality Subchannel thermal-hydraulics included included Fuel performance included included Coolant chemistry

  20. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    of plant operation and fuel rod design on the thermo-mechanical behavior, including Pellet-Cladding Interaction (PCI) failures in PWRs. The multi-physics, multi-dimensional...

  1. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    to deliver materials insight in the areas of CRUD, Grid-to-Rod-Fretting (GTRF), pellet-cladding interaction (PCI), reactivity insertion accident (RIA) and loss of cooling...

  2. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Reliable predictions of grid to rod gap, turbulent flow excitation, and resulting rod vibration and wear at any location in core. PCI Pellet-Clad Interaction. Cladding...

  3. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    well-known thermal-hydraulic analysis codes that have found widespread use in the nuclear energy industry. This group of codes is related in that they were developed for modeling...

  4. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    to develop the world's first nuclear fuel cycle and today is DOE's largest science and energy laboratory. ORNL has world-leading capabilities in computing and computational...

  5. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    plan to set up eight innovation hubs to solve the eight biggest energy problems in the world. CUNY Energy Institute The CUNY Energy Institute is proudly training the next...

  6. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Subchannel Methods for the Thermal-Hydraulic Analysis for Nuclear Power Systems presented by Dr. Michael Doster...

  7. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Monte Carlo Code for Research and Development," ... Roughness Elements," International Journal of Heat and Fluid Flow, ... Conference on Mathematics and Computation (M&C), ...

  8. The Consortium for Advanced Simulation of Light Water Reactors

    SciTech Connect

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  9. Tabular water properties interface for Hydra-TH : CASL THM.CFD.P6.03 milestone report.

    SciTech Connect

    Carpenter, John H.; Belcourt, Noel

    2013-04-01

    Completion of the CASL L3 milestone THM.CFD.P6.03 provides a tabular material properties capability to the Hydra code. A tabular interpolation package used in Sandia codes was modified to support the needs of multi-phase solvers in Hydra. Use of the interface is described. The package was released to Hydra under a government use license. A dummy physics was created in Hydra to prototype use of the interpolation routines. Finally, a test using the dummy physics verifies the correct behavior of the interpolation for a test water table. 3

  10. CASL-U-2015-0067-000 Virtual Environment

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    7-000 Virtual Environment for Reactor Applications (VERA) Workshop Session 2: Hands on Training Rose Montgomery The Tennessee Valley Authority April 1, 2015 CASL-U-2015-0067-000 1 Virtual Environment for Reactor Applications (VERA) Hands-On Training The American Nuclear Society ANFM Topical Meeting presents April 1, 2015 Hilton Head, SC Please check in to receive your student packet and RSA token to log onto the computers. CASL-U-2015-0067-000 2 2 CASL-U-2015-0067-000 2 CASL-U-2015-0067-000

  11. CASL-8-2015-0102-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    2-000 Lean/Agile Principles and Practices for Developing Quality Scientific Software Roscoe A. Bartlett Oak Ridge National Laboratory July 8-10, 2013 CASL-U-2015-0102-000 Lean/Agile Principles and Practices for Developing Quality Scientific Software Roscoe A. Bartlett CASL Vertical Reactor Integration Software Engineering Lead Trilinos Software Engineering Technologies and Integration Lead Computer Science and Mathematics Div CASL-U-2015-0102-000 2 Managed by UT-Battelle for the U.S. Department

  12. The CASL vision is to confidently predict

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL vision is to confidently predict the performance of commercial nuclear power reactors through comprehensive, science-based modeling and simulation technology. To achieve this vision, CASL is assembling, assessing and coupling a variety of phys- ics codes, each with a distinct purpose and functionality. This higher-fidelity coupled physics code capability is intended to have broad, versatile functionality with multiple modules simulating issues such as grid-to-rod-fretting and CRUD

  13. CASL Test Stands . . . Enabling the CASL mission

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    A comprehensive, science-based understanding of these complex challenges requires high fidelity modeling & simulation (modsim) technology. CASL brings together a...

  14. CASL Industry Council Meeting

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    ... DOE O 414.1D, ISO 9001-2008, and NQA-1-20082009 Part IV Subpart 4.2. The CASL ... independent review of the CASL QAP by an NQA-1 qualified lead auditor was recently ...

  15. CASL Milestone L2.VRI.P6.01

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    L2 Milestone VRI.P6.04 - VERA 3.0 CASL-U-2014-0006-000 i Fe Milestone L2:VRI.P6.04 Virtual Environment for Reactor Applications (VERA): Snapshot 3.0 03/29/2013 CASL-U-2014-0006-000 CASL L2 Milestone VRI.P6.04 - VERA 3.0 CASL-U-2014-0006-000 ii Oak Ridge National Laboratory in partnership with Electric Power Research Institute Idaho National Laboratory Los Alamos National Laboratory Massachusetts Institute of Technology North Carolina State University Sandia National Laboratories Tennessee Valley

  16. Light Water Reactor Sustainability (LWRS) Program | Department...

    Energy Saver

    Nuclear Reactor Technologies Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) ...

  17. Light Water Reactor Sustainability Technical Documents | Department...

    Energy Saver

    Nuclear Reactor Technologies Light Water Reactor Sustainability Program Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical ...

  18. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  19. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  20. CASL-U-2015-0172-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    2-000 3D Discrete Ordinates Reactor Assembly Calculations on GPUs Tom Evans Wayne Joubert, Steven Hamilton, Seth Johnson, John Turner, Greg Davidson, Tara Pandya Oak Ridge National Laboratory April 19, 2015 CASL-U-2015-0172-000 ORNL is managed by UT-Battelle for the US Department of Energy 3D Discrete Ordinates Reactor Assembly Calculations on GPUs Tom Evans Wayne Joubert Steven Hamilton Seth Johnson John Turner Greg Davidson Tara Pandya April 21, 2015 CASL-U-2015-0172-000 2 3D S N Problems on

  1. Microsoft PowerPoint - NEAC CASL @ 3.5 Years Report v4

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    ... Reactor System Multiphysics Integrator 4 CASL @3.5 Years Mesh Motion Quality Improvement Geometry Management Mesh Motion Quality Improvement Geometry Management Virtual ...

  2. CASL - Keeping It Real

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Keeping It Real To ensure its modeling and simulation research activities are appropriately focused on specific industry issues, CASL engages an Industry Council chaired by the Electric Power Research Institute. CASL is committed to engaging the nuclear power industry to ensure its research is "used and useful" and positively contributes to the ability to generate safe, reliable, economical, and carbon-free electricity from nuclear power. In his visit to the CASL facilities at Oak

  3. CASL - Westinghouse Electric Company

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Definition of CASL challenge problems Existing codes and expertise Data for validation Computatinoal fluid dynamics modeling and analysis Development of test stand for Virtual ...

  4. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL-U-2013-0105-000 Operational Reactor Model Demonstration with VERA: Watts Bar Unit 1 Cycle 1 Zero Power Physics Tests Jess C Gehin, ORNL Andrew T Godfrey, ORNL Fausto Franceschini, Westinghouse Advanced Modeling Applications Thomas M Evans, ORNL Benjamin S Collins, UM Steven P Hamilton, ORNL Radiation Transport Methods Revision 1 August 23, 2013 CASL-U-2013-0105-001 Operational Reactor Model Demonstration with VERA CASL-U-2013-0105-001 i Consortium for Advanced Simulation of LWRs Oak Ridge

  5. CASL - Analysis of Two-Dimensional Lattice Physics

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Analysis of Two-Dimensional Lattice Physics CASL is developing the Virtual Environment for Reactor Applications (VERA) as a key capability to support the analysis of the CASL Challenge Problems. VERA will include a range of physics modeling capabilities necessary to model reactors, including neutronics, thermal hydraulics, fuel performance, and coolant chemistry. Lattice physics analyses, utilizing the newly-developed Michigan lattice physics neutronics capability in MPACT 1.0, are important to

  6. CASL - Implementing Integrated Steam Tables

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Implementing Integrated Steam Tables At the end of August, CASL published an equation of state model library for use with VERA. The IAPWS95 and IAPWS-IF97 standard models for the thermodynamic properties of water and the associated transport property models implemented within the library were verified to reproduce the analytic models across the range of validity. The performance of the interpolation package is over an order of magnitude faster than the analytic model equations, even for tables

  7. CASL - Initial Modeling and Analysis of the Departure from Nucleate Boiling

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Challenge Problem Initial Modeling and Analysis of the Departure from Nucleate Boiling Challenge Problem Yixing Sung, Jin Yan, Zeses E. Karoutas of Westinghouse Electric Company LLC Anh V. Bui, Hongbin Zhang of Idaho National Laboratories Nam Dinh of North Carolina State University Departure from Nucleate Boiling (DNB) is one of the safety-related Challenge Problems (CP) that CASL is addressing in support of Pressurized Water Reactor (PWR) power uprate, high fuel burnup and plant lifetime

  8. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  9. CASL Industry Council Members:

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL Industry Council Members: We are looking forward to hosting you at the upcoming CASL Industry Council Meeting on Tuesday, April 12, 2016 through Wednesday, April 13, 2016 at the following location: ALOFT Greenville Downtown Converge Conference Room 5 North Laurens Street Greenville, SC 29601 864-297-6100 Meeting Contact: Lorie Fox (865) 548-5178 Lodging: ALOFT Greenville Downtown: http://www.aloftgreenvilledowntown.com/ Hotel Information * Check-in time: 4 PM * Checkout time: 12 PM * Fast

  10. CASL Test Stand Experience

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Industry Test Stand Experience Stephen Hess, EPRI Heather Feldman, EPRI Brenden Mervin, EPRI Martin Pytel, EPRI Rose Montgomery, TVA Bill Bird, TVA Fausto Franceschini, Westinghouse Electric Company LLC Advanced Modeling Applications 28 March 2014 CASL-U-2014-0036-000 Consortium for Advanced Simulation of LWRs ii CASL-U-2014-0036-000 REVISION LOG Revision Date Affected Pages Revision Description 0 3/28/2014 All Original Report Document pages that are: Export Controlled

  11. CASL-U-2015-0233-000 CASL Program Highlights

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    3-000 CASL Program Highlights January 2015 Douglas Kothe Oak Ridge National Laboratory January 31, 2015 CASL-U-2015-0233-000 Computational Science Efforts Advance Fuel Performance Simulation MOOSE/Bison-CASL capability more adept at simulating the important fuel pellet-cladding contact phenomena * A new mechanical contact system in MOOSE has resulted in significantly better solution convergence for Bison-CASL * Size of models that can be run with this system had previously been limited due to

  12. Light Water Reactor Sustainability Program - Integrated Program...

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    Light Water Reactor Sustainability Program - Integrated Program Plan Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability ...

  13. Development of a New 47-Group Library for the CASL Neutronics Simulators

    SciTech Connect

    Kim, Kang Seog; Williams, Mark L; Wiarda, Dorothea; Godfrey, Andrew T

    2015-01-01

    The CASL core simulator MPACT is under development for the neutronics and thermal-hydraulics coupled simulation for the pressurized light water reactors. The key characteristics of the MPACT code include a subgroup method for resonance self-shielding, and a whole core solver with a 1D/2D synthesis method. The ORNL AMPX/SCALE code packages have been significantly improved to support various intermediate resonance self-shielding approximations such as the subgroup and embedded self-shielding methods. New 47-group AMPX and MPACT libraries based on ENDF/B-VII.0 have been generated for the CASL core simulator MPACT of which group structure comes from the HELIOS library. The new 47-group MPACT library includes all nuclear data required for static and transient core simulations. This study discusses a detailed procedure to generate the 47-group AMPX and MPACT libraries and benchmark results for the VERA progression problems.

  14. CASL - Materials and Performance Optimization (MPO)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Materials and Performance Optimization (MPO) The Materials and Performance Optimization (MPO) focus area within CASL has recently developed and released a 3D modeling framework known as MAMBA (MPO Advanced Model for Boron Analysis) to predict CRUD deposition on nuclear fuel rods. CRUD, which refers to Chalk River Unidentified Deposit, is predominately a nickel-ferrite spinel corrosion product that deposits on hot fuel clad surfaces in nuclear reactors. CRUD has a lower thermal conductivity than

  15. WATER BOILER REACTOR

    DOEpatents

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  16. CASL Education Program and Summer Institute

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    0-000 CASL Education Program and Summer Institute Dr. J. Michael Doster Professor of Nuclear Engineering North Carolina State University Director, CASL Education Program 2 CASL-U-2016-1090-000 CASL Education Program A new generation of LWR Designers, Scientists, and Nuclear Power Professionals Program Charter: * Integrate CASL technology into undergraduate and graduate curricula * Develop a plan to educate industry users * Encourage diversity of participation in CASL activities * Advise Chair on

  17. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    L3:RTM.PRT.P9.02 Demonstration of Full Core Reactor Depletion with MPACT Revision 0 Benjamin Collins Aaron Graham Ang Zhu Brendan Kochunas Tom Downar Radiation Transport Methods Oak Ridge National Laboratory August 10, 2014 CASL-U-2014-0140-000 Demonstration of Full Core Reactor Depletion with MPACT Consortium for Advanced Simulation of LWRs i CASL-U-2014-0140-000 REVISION LOG Revision Date Affected Pages Revision Description 0 8/10/2014 All Original Report Document pages that are: Export

  18. CASL-U-2015-0230-000 CASL Program Highlights

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Plan * Uncertainty Quantification for MPACT * Industry Council Actions * Licensing Strategy Update * Education Program FY15 Plan * VERA Release Topics * CASL 2.0 Kickoff...

  19. CASL OLCF Early Science Award

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    OLCF Early Science Award CASL-U-2013-0231-000 i Fe High-Fidelity Neutronic Analysis of the Westinghouse AP1000 An Oak Ridge Leadership Computing Facility (OLCF) Early Science Demonstration using the GPU- accelerated Cray XK7 Titan System Tom Evans Fausto Franceschini Andrew Godfrey Steven Hamilton Wayne Joubert John Turner CASL-U-2013-0231-000 CASL OLCF Early Science Award CASL-U-2013-0231-000 ii Oak Ridge National Laboratory in partnership with Electric Power Research Institute Idaho National

  20. Light Water Reactor Sustainability Technical Documents | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high...

  1. CASL-U-2015-0234-000 CASL Program Highlights

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    4-000 CASL Program Highlights February 2015 Douglas Kothe Oak Ridge National Laboratory February 29, 2014 CASL-U-2015-0234-000 Outcome: VERA was demonstrated to be capable of simulating iPWR geometry and control rod management for normal operating conditions. H concentration Contributors: Rose Montgomery (TVA), Ben Collins (ORNL) Independent review by Dudley Raine (B&W) CASL Demonstration of VERA Capabilities to simulate an iPWR SMR Illustration of the B&W mPower iPWR used as a basis for

  2. Light Water Reactor Sustainability Program - Integrated Program...

    Energy Saver

    Program - Integrated Program Plan Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and ...

  3. Improving Reactor Performance Rose Montgomery The Tennessee Valley...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Improving Reactor Performance Rose Montgomery The Tennessee Valley Authority Mark Uhran Oak Ridge National Laboratory April 9, 2013 CASL-U-2013-0034-001 CASL-U-2013-0034-001 ...

  4. CASL-U-2015-0243-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    3-000 CFD Simulations of Experiments in a Twin Jet Water Facility Lane Carasik Yassin Hassan Texas A&M University CASL-U-2015-0243-000 CFD Simulations of Experiments in a Twin Jet Water Facility Lane Carasik, Graduate Research Assistant and NEUP Fellow Dr. Yassin Hassan, Professor Twin Jet Studies Turbulent twin jets shooting into quiescent surroundings provide a physics rich platform for developing and validating new simulations. The jet properties can be controlled in the TJWF built to

  5. CASL L1 Milestone report : CASL.P4.01, sensitivity and uncertainty analysis for CIPS with VIPRE-W and BOA.

    SciTech Connect

    Sung, Yixing; Adams, Brian M.; Secker, Jeffrey R.

    2011-12-01

    The CASL Level 1 Milestone CASL.P4.01, successfully completed in December 2011, aimed to 'conduct, using methodologies integrated into VERA, a detailed sensitivity analysis and uncertainty quantification of a crud-relevant problem with baseline VERA capabilities (ANC/VIPRE-W/BOA).' The VUQ focus area led this effort, in partnership with AMA, and with support from VRI. DAKOTA was coupled to existing VIPRE-W thermal-hydraulics and BOA crud/boron deposit simulations representing a pressurized water reactor (PWR) that previously experienced crud-induced power shift (CIPS). This work supports understanding of CIPS by exploring the sensitivity and uncertainty in BOA outputs with respect to uncertain operating and model parameters. This report summarizes work coupling the software tools, characterizing uncertainties, and analyzing the results of iterative sensitivity and uncertainty studies. These studies focused on sensitivity and uncertainty of CIPS indicators calculated by the current version of the BOA code used in the industry. Challenges with this kind of analysis are identified to inform follow-on research goals and VERA development targeting crud-related challenge problems.

  6. CHIMNEY FOR BOILING WATER REACTOR

    DOEpatents

    Petrick, M.

    1961-08-01

    A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

  7. CASL Validation Data: An Initial Review

    SciTech Connect

    Nam Dinh

    2011-01-01

    The study aims to establish a comprehensive view of “data” needed for supporting implementation of the Consortium of Advanced Simulation of LWRs (CASL). Insights from this review (and its continual refinement), together with other elements developed in CASL, should provide the foundation for developing the CASL Validation Data Plan (VDP). VDP is instrumental to the development and assessment of CASL simulation tools as predictive capability. Most importantly, to be useful for CASL, the VDP must be devised (and agreed upon by all participating stakeholders) with appropriate account for nature of nuclear engineering applications, the availability, types and quality of CASL-related data, and novelty of CASL goals and its approach to the selected challenge problems. The initial review (summarized on the January 2011 report version) discusses a broad range of methodological issues in data review and Validation Data Plan. Such a top-down emphasis in data review is both needed to see a big picture on CASL data and appropriate when the actual data are not available for detailed scrutiny. As the data become available later in 2011, a revision of data review (and regular update) should be performed. It is expected that the basic framework for review laid out in this report will help streamline the CASL data review in a way that most pertinent to CASL VDP.

  8. Microsoft PowerPoint - CASL IC member Overviews.pptx

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Meeting ALOFT Hotel, Greenville, SC April 12-13, 2016 CASL-U-2016-1092-000 2 2 Meeting Objectives Exchange of information about CASL's research and activities to: 1. Provide an opportunity for engagement between industry stakeholders and CASL researchers. 2. Present and seek feedback on the progress on CASL's R&D activities and plans. 3. Discuss CASL and industry priorities to ensure that they are aligned. 4. Identify strategic collaborations between industry and CASL Focus Areas. Industry

  9. HEAVY WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Szilard, L.

    1958-04-29

    A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

  10. Light Water Reactor Sustainability Nondestructive Evaluation...

    Energy Saver

    Nondestructive Evaluation for Concrete Research and Development Roadmap Light Water ... US Department of Energy Office of Nuclear Energy's Light Water Reactor Sustainability ...

  11. LIGHT WATER MODERATED NEUTRONIC REACTOR

    DOEpatents

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  12. CASL - Sandia National Laboratory

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Key Outcomes Virtual Environment for Reactor Application (VERA) Advanced computational fluid dynamics (CFD) capabilities Advanced structural mechanics capabilities Multi-physics ...

  13. CASL - Idaho National Laboratory

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    The laboratory has designed and operated 52 test reactors, including EBR-1, the world's first nuclear power plant Key Contributions System safety analysis Multiscale fuel ...

  14. CASL---U---2014-0038-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    User Guidelines and Best Practices for CASL VUQ Analysis Using Dakota CASL---U---2014-0038-000 Brian M. Adams 1 Russell W. Hooper 1 Allison Lewis 2 Jerry A. McMahan, Jr. 2 Ralph C. S mith 2 Laura P . Swiler 1 Brian J. Williams 3 1 Sandia N ational L aboratories 2 North C arolina S tate U niversity 3 Los A lamos National Laboratory March 2 8, 2014 L3:VUQ.V&V.P8.01 CASL-U-2014-0038-000 User Guidelines and Best Practices for CASL VUQ Analysis Using Dakota Brian M. Adams 1 Russell W. Hooper 2

  15. CASL-8-2015-0109-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    9-000 Finding the Cure for CRUD: Combined Simulations & Experiments at CASL & MIT Michael P. Short Massachusetts Institute of Technology July 9, 2013 CASL-U-2015-0109-000 CASL Summer Student Workshop Oak Ridge National Laboratory 2013-07-09 Finding the Cure for CRUD: Combined Simulations & Experiments at CASL & MIT Prof. Michael P. Short (MIT) With Thanks To: D. Hussey (EPRI) D. Gaston, C. Permann (INL) D. Andersson, B. Kendrick, C. Stanek (LANL) I. Dumnernchanvanit, A. Dykhuis,

  16. CASL-8-2015-0220-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    220-000 Update on Anhomonic Entropic Effects on Zr-H System Thermodynamics Christopher Stanek Los Alamos National Laboratory June 3, 2015 CASL-U-2015-0220-000 CASL-U-2015-0220-000 Starting point should be a comprehensive thermodynamic description 1 CASL-U-2015-0220-000 Oxide microstructure ZrO 2 ZrO Zr 2 O Zr(O) Zr(O) Zr 2 O ZrO ZrO 2 Courtesy of Emmanuelle Marquis 2 CASL-U-2015-0220-000 composition strain Coupled chemistry/mechanics free energy description Interface free energies Atomic

  17. CASL-U-2015-0036-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    MC2015 - Joint International Conference on Mathematics and Computation ... CASL (Consortium for Advanced Simulation of LWRs) ... PN equations," Journal of Computational ...

  18. CASL-8-2015-0021-000 Contact Memory Report CASL.FY14 Rich Williamson

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    1-000 Contact Memory Report CASL.FY14 Rich Williamson Idaho National Laboratory January 31, 2015 CASL-U-2015-0021-000 Contact Memory Reduction CASL FY15 Letter Report January 30, 2015 Introduction As documented in a prior CASL milestone report [1], work has been underway to transition to a new system of contact enforcement in MOOSE, and by extension, the fuel performance codes BISON and BISON-CASL, which are based on MOOSE. This new system, known as the Constraint system, achieves significantly

  19. SUPERHEATING IN A BOILING WATER REACTOR

    DOEpatents

    Treshow, M.

    1960-05-31

    A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

  20. Heavy Water Test Reactor Dome Removal

    SciTech Connect

    2011-01-01

    A high speed look at the removal of the Heavy Water Test Reactor Dome Removal. A project sponsored by the Recovery Act on the Savannah River Site.

  1. PRESSURIZED WATER REACTOR CORE WITH PLUTONIUM BURNUP

    DOEpatents

    Puechl, K.H.

    1963-09-24

    A pressurized water reactor is described having a core containing Pu/sup 240/ in which the effective microscopic neutronabsorption cross section of Pu/sup 240/ in unconverted condition decreases as the time of operation of the reactor increases, in order to compensate for loss of reactivity resulting from fission product buildup during reactor operation. This means serves to improve the efficiency of the reactor operation by reducing power losses resulting from control rods and burnable poisons. (AEC)

  2. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    Office of Energy Efficiency and Renewable Energy (EERE)

    Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary November 2014

  3. CASL L3 Report: Tally Server implementation in OpenMC

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    MCH.P6.03 Tally Server Implementation in OpenMC Paul K. Romano, B. Forget MIT October 24, 2013 CASL-U-2013-0215-000 CASL L3 Report: Tally Server implementation in OpenMC Paul K. Romano and B.Forget October 24, 2013 Abstract As part of POR-6, an L3 milestone was completed on the implementation of tally servers in OpenMC. In the present work, we make a number of refinements to a theoretical performance model of the tally server algorithm to better predict the performance of a realistic reactor

  4. Light-Water Breeder Reactor

    DOEpatents

    Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  5. Specification of requirements for the virtual environment for reactor applications simulation environment

    SciTech Connect

    Hess, S. M.; Pytel, M.

    2012-07-01

    In 2010, the United States Dept. of Energy initiated a research and development effort to develop modern modeling and simulation methods that could utilize high performance computing capabilities to address issues important to nuclear power plant operation, safety and sustainability. To respond to this need, a consortium of national laboratories, academic institutions and industry partners (the Consortium for Advanced Simulation of Light Water Reactors - CASL) was formed to develop an integrated Virtual Environment for Reactor Applications (VERA) modeling and simulation capability. A critical element for the success of the CASL research and development effort was the development of an integrated set of overarching requirements that provides guidance in the planning, development, and management of the VERA modeling and simulation software. These requirements also provide a mechanism from which the needs of a broad array of external CASL stakeholders (e.g. reactor / fuel vendors, plant owner / operators, regulatory personnel, etc.) can be identified and integrated into the VERA development plans. This paper presents an overview of the initial set of requirements contained within the VERA Requirements Document (VRD) that currently is being used to govern development of the VERA software within the CASL program. The complex interdisciplinary nature of these requirements together with a multi-physics coupling approach to realize a core simulator capability pose a challenge to how the VRD should be derived and subsequently revised to accommodate the needs of different stakeholders. Thus, the VRD is viewed as an evolving document that will be updated periodically to reflect the changing needs of identified CASL stakeholders and lessons learned during the progress of the CASL modeling and simulation program. (authors)

  6. CASL-2015.L2.FMC.P11.01

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    &$6/ 8 L2.FMC.P11.01 Demonstrate 3D PCI analysis with BISON- CASL for PCI fuel failure in a commercial power reactor Nathan C apps 1 , W enfeng L iu 2 , R obert Montgomery 3 , M artin P ytel 4 , J oe R ashid 2 , Ben Spencer 5 , D ion S underland 3 , R ich Williamson 5 , a nd B rian D . W irth 1 1. University of Tennessee, Knoxville 2. Anatech Corp. 3. Pacific Northwest National Laboratory 4. Electric Power Research Institute 5. Idaho National Laboratory September 30, 2015

  7. CASL-U-2015-0167-000 COBRA-TF

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    7-000 COBRA-TF Parallelization and Application to PWR Reactor Core Vefa Kucukboyaci and Yixing Sung Westinghouse Electric Company Robert Salko Oak Ridge National Laboratory April 19, 2015 CASL-U-2015-0167-000 ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, TN * April 19-23, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) COBRA-TF PARALLELIZATION AND

  8. TA-2 Water Boiler Reactor Decommissioning Project

    SciTech Connect

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

  9. CASL-U-2015-0152-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    2-000 Simulation of CASL 3D HFP Fuel Assembly Benchmark Problem with On-The-Fly Doppler Broadening in MCNP6 Scott J. Wilderman and William R. Martin University of Michigan Forrest B. Brown Los Alamos National Laboratory March 29, 2015 CASL-U-2015-0152-000 Advances in Nuclear Fuel Management V (ANFM 2015) Hilton Head Island, South Carolina, USA, March 29 - April 1, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) © ANS 2015, Topical Meeting ANFM 2015, p. 1/7 SIMULATION OF CASL

  10. Review of light water reactor safety

    SciTech Connect

    Cheng, H.S.

    1980-12-01

    A review of the present status of light water reactor (LWR) safety is presented. The review starts with a brief discussion of the outstanding accident scenarios concerning LWRs. Where possible the areas of present technological uncertainties are stressed. To provide a better perspective of reactor safety, it then reviews the probabilistic assessment of the outstanding LWR accidents considered in the Reactor Safety Study (WASH-1400) and discusses the potential impact of the present technological uncertainties on WASH-1400.

  11. Boiling water reactor-full length emergency core cooling heat...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Boiling water reactor-full length emergency core cooling heat transfer ... Citation Details In-Document Search Title: Boiling water reactor-full length emergency ...

  12. Preliminary design study of small long life boiling water reactor...

    Office of Scientific and Technical Information (OSTI)

    boiling water reactor (BWR) with tight lattice thorium nitride fuel Citation Details In-Document Search Title: Preliminary design study of small long life boiling water reactor ...

  13. Partial Defect Testing of Pressurized Water Reactor Spent Fuel...

    Office of Scientific and Technical Information (OSTI)

    Partial Defect Testing of Pressurized Water Reactor Spent Fuel Assemblies Citation Details In-Document Search Title: Partial Defect Testing of Pressurized Water Reactor Spent Fuel ...

  14. Accident analysis of heavy water cooled thorium breeder reactor...

    Office of Scientific and Technical Information (OSTI)

    Accident analysis of heavy water cooled thorium breeder reactor Citation Details In-Document Search Title: Accident analysis of heavy water cooled thorium breeder reactor ...

  15. Development of Light Water Reactor Fuels with Enhanced Accident...

    Energy Saver

    Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to Congress Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to ...

  16. Light Water Reactor Sustainability Program - Non-Destructive...

    Energy Saver

    Light Water Reactor Sustainability Program - Non-Destructive Evaluation R&D Roadmap for ... important information to the Light Water Reactor Sustainability (LWRS) program ...

  17. Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap...

    Energy Saver

    Damage in Piping Light Water Reactor Sustainability (LWRS) Program - R&D Roadmap for Non-Destructive Evaluation (NDE) of Fatigue Damage in Piping Light water reactor sustainability ...

  18. CASL-U-2015-0223-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    W. Larsen University of Michigan May 28, 2015 Stability of Monte Carlo k-Eigenvalue ... Journal of Computational Physics May 28, 2015 CASL-U-2015-0223-000 with the coarse grid ...

  19. CASL-U-2015-0251-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Heat Transfer Regimes in CRUD Miaomiao Jin, Michael P. Short Massachusetts Institute of Technology July 7, 2015 CASL-U-2015-0251-000 STUDY ON TWO-PHASE FLUID AND HEAT TRANSFER ...

  20. Phase 2 CASL Unlimited Access Report Outline

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    www.casl.gov CASL-U-2015-0189-000 Evaluation of Missing Pellet Surface Geometry on Cladding Stress Distribution and Magnitude Nathan Capps and Brian D. Wirth University of Tennessee Robert Montgomery and Dion Sunderland Pacific Northwest National Laboratory Benjamin Spencer Idaho National Laboratory Martin Pytel Electric Power Research Institute May 1, 2015 Evaluation of Missing Pellet Surface Geometry on Cladding Stress Distribution and Magnitude Babcock & Wilcox Critical Consortium for

  1. CASL-8-2015-0026-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    6-000 Fuel Performance with BISON Koroush Shirvan Massachusetts Institute of Technology July 12, 2014 CASL-U-2015-0026-000 Koroush Shirvan Research Scientist June 12, 2014 Fuel performance with BISON NSE Nuclear Science & Engineering at MIT science : systems : society CASL-U-2015-0026-000 Outline  BISON Overview  Prerequisite  Fuel Performance Overview  Oxide Fuel Pin Behavior  Reactivity Initiating Accidents  Metal Fuel Pin Behavior  BISON Demonstration  Conclusions

  2. CASL-U-2015-0022-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    2-000 Incorporate MPACT into TIAMAT and Demonstrate Pellet-Clad Interaction (PCI) Calculations Kevin Clarno Oak Ridge National Laboratory Roger Pawlowski Sandia National Laboratory December 16, 2014 CASL-U-2015-0022-000 K. Clarno and R. Pawlowski ABSTRACT CASL (the Consortium for Advanced Simulation of LWRs) is investing in the development of the multi-dimensional Bison fuel performance code for both transient and nominal peformance analysis. The MPACT whole-core neutronics capability has been

  3. Light Water Reactor Sustainability (LWRS) Initiative Science...

    Energy Saver

    disposed instead of untreated used fuel. April 29, 2010 Constituents of Used Light Water Reactor Nuclear Fuel (by mass) April 29, 2010 Descriptions from NE R&D Roadmap to...

  4. CASL - Radiation Transport Methods Update

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Radiation Transport Methods Update The Radiation Transport Methods (RTM) focus area is responsible for the development of methods, algorithms, and implementations of radiation transport methods as they apply to the design and analysis of light water nuclear reactors. the fundamental areas of investigation in RTM include high-order deterministic transport low-order transport approximations multigroup cross section generation depletion as it applies to in-core neutronics and material coupling

  5. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Review of Experiments for CASL Neutronics Validation HONGBIN ZHANG Idaho National Laboratory March 29, 2012 CASL-U-2012-0039-000 Review of Experiments for CASL Neutronics Validation ii CASL-U-2012-0039-000 Please complete sections appropriate for this record. REVISION LOG Revision Date Affected Pages Revision Description N/A Initial version Document pages that are: Export Controlled __________________________________________________ IP/Proprietary/NDA

  6. Pressurized water reactor flow skirt apparatus

    DOEpatents

    Kielb, John F.; Schwirian, Richard E.; Lee, Naugab E.; Forsyth, David R.

    2016-04-05

    A pressurized water reactor vessel having a flow skirt formed from a perforated cylinder structure supported in the lower reactor vessel head at the outlet of the downcomer annulus, that channels the coolant flow through flow holes in the wall of the cylinder structure. The flow skirt is supported at a plurality of circumferentially spaced locations on the lower reactor vessel head that are not equally spaced or vertically aligned with the core barrel attachment points, and the flow skirt employs a unique arrangement of hole patterns that assure a substantially balanced pressure and flow of the coolant over the entire underside of the lower core support plate.

  7. Materials Degradation in Light Water Reactors: Life After 60

    Office of Energy Efficiency and Renewable Energy (EERE)

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field....

  8. CASL - VERA-CS Coupled Multi-physics Capability demonstrated in a Full Core

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Simulation VERA-CS Coupled Multi-physics Capability demonstrated in a Full Core Simulation In December, CASL reported on the latest results from its Watts Bar reactor progression problem modeling. The most recent simulation includes quarter-core representation with coupled neutronics (including embedded cross section generation and neutron transport) and thermal - hydraulics. Last July, the team completed a demonstration of physics coupling that included heat generation and thermal-hydraulic

  9. CASL-8-2015-0199-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    99-000 CASL Meeting with the Illinois Applied Research Institute Douglas Kothe and Jess Gehin at the University of Illinois at Urbana- Champaign October 6, 2014 CASL-U-2015-0199-000 1 Meeting Agenda Visit of Dr. Douglas Kothe and Dr. Jess Gehin at the University of Illinois Monday, October 6 th , 2014 9:00 am - 4:00 pm Discussion on a pathway to build a Test Stand in collaboration with the Oak Ridge National Laboratory and the University of Illinois. Specific interest: training and education,

  10. CASL-U-2015-0116-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    16-000 Implementation and Preliminary Verification of the Quasi-1D Kernel for Intra-pin Resonance Physics in MPACT Yuxuan Liu and William Martin University of Michigan March 16, 2015 CASL-U-2015-0116-000 Implementation and Preliminary Verification of the Quasi-1D Kernel for Intra-pin Resonance Physics in MPACT Yuxuan Liu and William Martin University of Michigan L3:RTM.SUP.P10.03 Milestone Report March 16, 2015 CASL-U-2015-0116-000 2 EXECUTIVE SUMMARY A new resonance self-shielding method ESSM-X

  11. SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

    DOEpatents

    Ransohoff, J.A.; Plawchan, J.D.

    1960-08-16

    A boiling-water reactor was designed which comprises a pressure vessel containing a mass of water, a reactor core submerged within the water, a reflector tank disposed within the reactor, the reflector tank being open at the top to the interior of the pressure vessel, and a surge tank connected to the reflector tank. In operation the reflector level changes as a function of the pressure witoin the reactor so that the reactivity of the reactor is automatically controlled.

  12. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect

    Austin, W.; Brinkley, D.

    2011-10-13

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment

  13. CASL-U-2015-0108-000 A CASL Multiphysics Code Coupling Primer: Software

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    08-000 A CASL Multiphysics Code Coupling Primer: Software Integration 101 (SAND2013-5908C) Roger Pawlowski Sandia National Laboratory July 9, 2013 A CASL Multiphysics Code Coupling Primer: Software Integration 101 Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. Roger

  14. Hydrogen and water reactor safety: proceedings

    SciTech Connect

    Not Available

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  15. Light-water reactor accident classification

    SciTech Connect

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art.

  16. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    DOEpatents

    Treshow, M.

    1963-04-30

    This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

  17. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  18. CASL - Los Alamos National Laboratory

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Accurate materials models and data, integrated into the Virtual Reactor (VR) simulation tool for science-based prediction Advanced thermomechanical, fluid dynamics, neutron ...

  19. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Evaluation of Industry Council Pilot Project Alternatives (AMA.REQ.P4.01) 29 February 2012 CASL-U-2012-0026-000 Evaluation of Industry Council Pilot Project Alternatives ii CASL-U-2012-0026-000 REVISION LOG Revision Date Affected Pages Revision Description 0.0 02/29/12 All New issue. Evaluation of Industry Council Pilot Project Alternatives iii CASL-U-2012-0026-000 CONTENTS CONTENTS

  20. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    VERA Core Physics Benchmark Progression Problem Specifications Revision 2 March 29, 2013 Andrew T. Godfrey Advanced Modeling Applications Oak Ridge National Laboratory A. Godfrey, VERA Core Physics Benchmark Progression Problem Specifications, CASL Technical Report: CASL-U-2012-0131-002 VERA Core Physics Benchmark Problems CASL-U-2012-0131-002 i Consortium for Advanced Simulation of LWRs This report was prepared as an account of work sponsored by an agency of the United States government.

  1. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe

    1994-01-01

    A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

  2. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, P.

    1994-07-05

    A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

  3. Light Water Reactor Sustainability Program: Materials Aging and...

    Energy Saver

    Program: Materials Aging and Degradation Technical Program Plan Light Water Reactor ... Primary water stress corrosion cracking (PWSCC) is one key form of degradation in extended ...

  4. CASL-U-2015-0056-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    a reduced set of user-input data. The utility produces models geared towards Pressurized-Water Reactor (PWR) rod-bundle geometry. It does this by using basic characteristics of...

  5. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL-U-2013-0273-000 Demonstration of Advanced Pin-Resolved MOC ... Distribution: To: NA Copy: NA Demonstration of Advanced Pin-Resolved MOC ...

  6. Five Years of Building the Next Generation of Reactors | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Energy Years of Building the Next Generation of Reactors Five Years of Building the Next Generation of Reactors August 15, 2012 - 5:17pm Addthis Simulated three-dimensional fission power distribution of a single 17x17 rod PWR fuel assembly. | Photo courtesy of the Consortium for Advanced Simulation of Light Water Reactors (CASL). Simulated three-dimensional fission power distribution of a single 17x17 rod PWR fuel assembly. | Photo courtesy of the Consortium for Advanced Simulation of Light

  7. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials

    Office of Energy Efficiency and Renewable Energy (EERE)

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the...

  8. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, C.W.

    1985-02-19

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  9. The 100K West Reactor Water Treatment Facilities

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    demolition (D&D) work at the 100K West Reactor Water Treatment Facilities at the Hanford ... facilities and waste sites that supported reactor operations from the 1950s to the 1970s. ...

  10. Development of Materials for Supercritical-Water-Cooled Reactor

    Energy.gov [DOE]

    Supercritical-Water-Cooled Reactor (SCWR) was selected as one of the promising candidates in Generation IV reactors for its prominent advantages; those are the high thermal efficiency, the system...

  11. Phase 2 CASL Unlimited Access Report Outline

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    34-000 MPACT Verification and Validation: Status and Plans Tom Downar Brendan Kochunas Univ of Michigan Ben Collins Oak Ridge National Laboratory April 30, 2015 MPACT Verification and Validation Consortium for Advanced Simulation of LWRs ii CASL-U-2015-0134-000 REVISION LOG Revision Date Affected Pages Revision Description 0 All Initial Release Document pages that are: Export Controlled ______________None________________________________ IP/Proprietary/NDA

  12. Phase 2 CASL Unlimited Access Report Outline

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    43-000 Babcock & Wilcox Critical Experiment Calculations to Support MPACT Verification J. A. Kulesza, University of Michigan S. G. Stimpson, Oak Ridge National Laboratory A. R. Gerlach, University of Michigan D. R. Jabaay, University of Michigan April 24, 2015 Babcock & Wilcox Critical Experiment Calculations to Support MPACT Verification Consortium for Advanced Simulation of LWRs ii CASL-U-2015-0143-000 REVISION LOG Revision Date Affected Pages Revision Description 0 All Initial Release

  13. Phase 2 CASL Unlimited Access Report Outline

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    5-0281-001 MPACT Library Verification by Comparison of Pincell Calculations to Monte Carlo Results Scott Palmtag Core Physics Inc. February 10, 2016 MPACT Pincell Calculations Consortium for Advanced Simulation of LWRs ii CASL-U-2015-0281-000 REVISION LOG Revision Date Affected Pages Revision Description 0 7/14/2015 All Initial Release 1 02/05/2016 All Major update Document pages that are: Export Controlled ________None______________________________________ IP/Proprietary/NDA

  14. Phase 2 CASL Unlimited Access Report Outline

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    15-000 Investigation of Thermal Expansion Effects in MPACT Scott Palmtag Core Physics Inc. February 28, 2016 MPACT Thermal Expansion Consortium for Advanced Simulation of LWRs ii CASL-U-2016-1015-000 REVISION LOG Revision Date Affected Pages Revision Description 0 2/28/2016 All Initial Release Document pages that are: Export Controlled ________None______________________________________ IP/Proprietary/NDA Controlled_________None_____________________________________ Sensitive

  15. Phase 2 CASL Unlimited Access Report Outline

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    52-000 MPACT Library Verification by Comparison of Assembly Calculations to Monte Carlo Results Scott Palmtag Core Physics Inc. March 31, 2016 MPACT Assembly Calculations Consortium for Advanced Simulation of LWRs ii CASL-U-2016-1052-000 REVISION LOG Revision Date Affected Pages Revision Description 0 03/31/2016 All Initial Release Document pages that are: Export Controlled ________None______________________________________ IP/Proprietary/NDA

  16. ValUQWorkflowInCASL-RiderFinal

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    0-0023-000 L3:VUQ.VVDA.P1-1.04 William Rider SNL Completed: 12/30/10 1 SAND2010-234P Unlimited Release December 2010 Verification, Validation and Uncertainty Quantification Workflow in CASL William J. Rider Computational Shock and MultiPhysics Department James R. Kamm and V. Gregory Weirs Optimization and Uncertainty Quantification Department Sandia National Laboratories P.O. Box 5800 Albuquerque, New Mexico 87185 Abstract The overall conduct of verification, validation and uncertainty

  17. CASL-U-2015-0020-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    0-000 Improvement of COBRA- TF Subchannel Thermal- Hydraulics Code (CTF) using Computational Fluid Dynamics Taylor S. Blyth Pennsylvania State University December 15, 2014 CASL-U-2015-0020-000 1 Improvement of COBRA-TF Subchannel Thermal-Hydraulics Code (CTF) using Computational Fluid Dynamics Taylor Blyth, The Pennsylvania State University, tyb5095@psu.edu 15 December 2014 Abstract This report presents a proposal for improving the CTF subchannel thermal hydraulics code using computational fluid

  18. CASL-U-2015-0252-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    2-000 Development of Advanced Analysis Tools for Interface Tracking Simulations Jun Fang, Igor A Bolotnov North Carolina State University July 7, 2015 CASL-U-2015-0252-000 Development of Advanced Analysis Tools for Interface Tracking Simulations Jun Fang, Dr. Igor A. Bolotnov Department of Nuclear Engineering, North Carolina State University Measurement of Lift and Drag Forces Subchannel Simulations and Computational Tools Development In order to evaluate the lift and drag forces, external

  19. CASL-U-2015-0253-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    3-000 Treatment of Nucleation and Bubble Dynamics in High Heat Flux Boiling Yang Liu, Nam Dinh North Carolina State University July 7, 2015 CASL-U-2015-0253-000 TREATMENT OF NUCLEATION AND BUBBLE DYNAMICS IN HIGH HEAT FLUX BOILING Yang Liu, Department of Nuclear Engineering, North Carolina State University Instructor: Dr. Nam Dinh, Department of Nuclear Engineering, North Carolina State University Nucleate boiling is a highly efficient and desirable cooling mechanism in high- power-density

  20. CASL-U-2015-0255-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    5-000 A Framework for Predictive Capability Maturity Quantification Paridhi Athe, Nam Dinh North Carolina State University July 7, 2015 CASL-U-2015-0255-000 A framework for Predictive Capability Maturity Quantification Paridhi Athe, Department of Nuclear Engineering, North Carolina State University Dr. Nam Dinh, Department of Nuclear Engineering, North Carolina State University Modeling and simulation permeate all fields of science and engineering. In nuclear engineering, numerical results

  1. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, Charles W.

    1987-01-01

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  2. (Boiling water reactor (BWR) CORA experiments)

    SciTech Connect

    Ott, L.J.

    1990-10-16

    To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of the BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.

  3. Environmentally assisted cracking in light water reactors

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  4. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    SciTech Connect

    T. R. Allen and G. S. Was

    2008-12-12

    Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept.

  5. Screening reactor steam/water piping systems for water hammer

    SciTech Connect

    Griffith, P.

    1997-09-01

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made.

  6. State space modeling of reactor core in a pressurized water reactor

    SciTech Connect

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W.; Shamsuddin, Mustaffa; Abdullah, M. Adib

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  7. Water inventory management in condenser pool of boiling water reactor

    DOEpatents

    Gluntz, Douglas M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  8. Water inventory management in condenser pool of boiling water reactor

    DOEpatents

    Gluntz, D.M.

    1996-03-12

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs.

  9. Code System for Supercritical Water Cooled Reactor LOCA Analysis.

    Energy Science and Technology Software Center

    1999-10-13

    Version 00 The new SCRELA code was developed to analyze the LOCA of the supercritical water cooled reactor. Since the currently available LWR codes for LOCA analysis could not analyze the significant differences in reactor characteristics between the supercritical-water cooled reactor and the current LWR, the first objective of this code development was to analyze the uniqueness of this reactor. The behavior of the supercritical water in the blowdown phase and the reflood phase ismore » modeled.« less

  10. Commercial Light Water Reactor Tritium Extraction Facility

    SciTech Connect

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  11. CASL-U-2014-0069-001

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    4-0069-001 Modeling an iPWR Startup Core Cycle with VERA Progress Report 2 Milestone L4:AMA.APP.P10.04 Kelly Kenner University of Tennessee - Knoxville Advisors: Ivan Maldonado, UTK Rose Montgomery, TVA Dudley Raine, B&W June 26, 2014 CASL-U-2014-0069-001 Page i Executive Summary The purpose of this report is to provide an update on the research work performed by Kelly Kenner, a Master's Candidate at the University of Tennessee's Department of Nuclear Engineering, as funded by Tennessee

  12. CASL-U-2015-0016-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    6-000 Advanced Calibration and Validation of a Mechanistic Model of Subcooled Boiling Two-Phase Flow Anh Bui, Idaho National Laboratory Brian Williams, Los Alamos National Laboratory Nam Dinh, North Carolina State University April 6, 2014 CASL-U-2015-0016-000 Proceedings of ICAPP 2014 Charlotte, USA, April 6-9, 2014 Paper 14257 Advanced Calibration and Validation of a Mechanistic Model of Subcooled Boiling Two-Phase Flow Anh Bui 1 , Brian Williams 2 , Nam Dinh 3,* 1 Idaho National Laboratory

  13. CASL-U-2015-0101-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    1-000 Sensitivity and Uncertainty Methods (SAND2013-5431C) Brian Adams Sandia National Laboratory July 8-10, 2013 CASL-U-2015-0101-000 Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. SAND2013-5431C based primarily on SAND2012-7387P, SAND2012-7388P Sensitivity and Uncertainty

  14. CASL-U-2015-0158-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    8-000 Transient Methods for Pin-Resolved Whole Core Transport Using the 2D-1D Methodology in MPACT Ang Zhu, Yunlin Xu, Aaron Graham, Mitchell Young, and Thomas Downar University of Michigan Liangzhi Cao Xi'an Jiaotong University April 19, 2015 CASL-U-2015-0158-000 ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, TN * April 19-23, 2015, on CD-ROM, American Nuclear

  15. CASL-U-2015-0165-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    5-000 The SCALE 6.2 ORIGEN API for High Performance Depletion W. A. Wieselquist Oak Ridge National Laboratory April 19, 2015 CASL-U-2015-0165-000 ANS MC2015-Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, TN * April 19-23, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) THE SCALE 6.2 ORIGEN API FOR HIGH PERFORMANCE DEPLETION W. A. Wieselquist Oak Ridge National

  16. CASL-U-2015-0176-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    6-000 Stability of S N K- Eigenvalue Iterations using CMFD Acceleration Kendra P. Keady and Edward W. Larsen University of Michigan April 19, 2015 CASL-U-2015-0176-000 ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, Tennessee * April 19-23, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) STABILITY OF S N K-EIGENVALUE ITERATIONS USING CMFD

  17. CASL-U-2015-0224-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    4-000 Stability of S N K= Eigenvalue Iterations using CMFD Acceleration Kendra P. Keady and Edward W. Larsen University of Michigan May 28, 2015 CASL-U-2015-0224-000 ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, Tennessee * April 19-23, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) STABILITY OF S N K-EIGENVALUE ITERATIONS USING CMFD ACCELERATION

  18. CASL-U-2015-0250-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    0-000 Development of a MCNP6 - ANSYS FLUENT Multiphysics Coupling Capability William Gurecky The University of Texas at Austin July 7, 2015 CASL-U-2015-0250-000 Development of a MCNP6 - ANSYS FLUENT Multiphysics Coupling Capability In this work geometries typical of operational U.S. PWRs are investigated to demonstrate a novel core physics coupling methodology. Monte Carlo based radiation transport (via MCNP6) and finite volume TH methodologies (via ANSYS-Fluent) are combined to achieve a

  19. CASL-U-2015-0254-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    4-000 Erosion of a Stratified Layer by a Buoyant Jet in a Large Vessel Fatih S. Sarikurt, Yassin Hassam Texas A&M University July 7, 2015 CASL-U-2015-0254-000 Erosion of a Stratified Layer by a Buoyant Jet in a Large Vessel Fatih Sinan Sarikurt, Graduate Research Assistant Dr. Yassin Hassan, Department Head and Professor of Nuclear Engineering Dimensions and Geometry The fine mesh was a polyhedral mesh with a 20 million cell count. Refinement areas were created and applied as required areas

  20. CASL-U-2015-0258-000

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    8-000 Sensitivity Study of MOC Parameters with VERA Benchmark Cases Jipu Wang, Bill Martin University of Michigan Benjamin Collins Oak Ridge National Laboratory July 7, 2015 CASL-U-2015-0258-000 Sensitivity Study of MOC Parameters with VERA Benchmark Cases Jipu Wang 1 (jipuwang@umich.edu) , Benjamin Collins 1,2 (collinsbs@ornl.gov), Bill Martin 1 (wrm@umich.edu) 1 Department of Nuclear Engineering and Radiological Sciences University of Michigan, Ann Arbor, MI 48109 2 Oak Ridge National

  1. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    1 Thermal Hydraulic Validation Report CASL Mile Stone: L3:AMA.VAL.P4.02 Authors: J. Yan, P. Yuan, Y. Xu, Zeses Karoutas Westinghouse Electric Company Emilio Popov, Sreekanth Pannala, Alan Stagg Oak Ridge National Lab Scott Lucas, Hongbin Zhang Idaho National Lab Robert Brewster, Emilio Baglietto CD-Adapco Elvis Dominguez-Ontiveros, Yassin Hassan Texas A&M March 2012 CASL-8-2012-0037-000 <Document Name> ii CASL-8-2012-0037-000 Please complete sections appropriate for this record. a)

  2. A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components

    Energy.gov [DOE]

    In the United States currently there are approximately 104 operating light water reactors. Of these, 69 are pressurized water reactors (PWRs) and 35 are boiling water reactors (BWRs). In 2007, the...

  3. CASL Milestone L2.AMA.P7.02

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    of LWRs Fe CASL-U-2013-0196-000 Demonstration of Neutronics Coupled to ... of Milestone L2:AMA.P7.02 - Demonstration of Neutronics Coupled to ...

  4. CASL-U-2015-0083-000 VERA Configure,

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    83-000 VERA Configure, Build, Test, and Install Quick Reference Guide Roscoe A. Bartlett Oak Ridge National Laboratory April 17, 2015 CASL-U-2015-0083-000 VERA Configure, Build,...

  5. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Analysis of Two-Dimensional Lattice Physics Verification Problems with MPACT Andrew Godfrey, ORNL Fausto Franceschini, Westinghouse Electric Company LLC Scott Palmtag, Core Physics Julie Stout, ORNL Advanced Modeling Applications December 21, 2012 CASL-U-2012-0172-000 Analysis of 2D Lattice Physics Verification Problems with MPACT CASL-U-2012-0172-000 ii Consortium for Advanced Simulation of LWRs REVISION LOG Revision Date Affected Pages Revision Description 0 12/21/2012 All Original Issue

  6. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    MPACT Testing and Benchmarking Results March 31, 2014 Andrew T. Godfrey Physics Integration Oak Ridge National Laboratory CASL-U-2014-0045-000 MPACT Testing and Benchmarking Consortium for Advanced Simulation of LWRs ii CASL-U-2014-0045-000 REVISION LOG Revision Date Affected Pages Revision Description 0 3/31/2014 All Original Release Document pages that are: Export Controlled __None_______________________________________________ IP/Proprietary/NDA

  7. CASL-U-2015-0175-000 VPSC

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    5-000 VPSC Implementation in BISON-CASL Code for Modeling Large Deformation Problems Wenfeng Liu ANATECH Corporation Robert Montgomery Pacific Northwest National Laboratory Carlos Tomé and Chris Stanek Los Alamos National Laboratory Jason Hales Idaho National Laboratory April 19, 2015 CASL-U-2015-0175-000 ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, TN * April

  8. CASL-U-2016-1030-000 Development

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Consortium for Advanced Simulation of LWRs CASL-U-2016-1030-000 Development and Testing of CTF to Support Modeling of BWR Operating Conditions Robert Salko, Oak Ridge National Laboratory Aaron Wysocki, Oak Ridge National Laboratory Benjamin Collins, Oak Ridge National Laboratory Andrew Godfrey, Oak Ridge National Laboratory Chris Gosdin, Pennsylvania State University Maria Avramova, North Carolina State University 01/29/2016 CASL-U-2016-1030-000 Page ii REVISION LOG Revision Date Affected Pages

  9. Process Intensification with Integrated Water-Gas-Shift Membrane Reactor |

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Department of Energy Intensification with Integrated Water-Gas-Shift Membrane Reactor Process Intensification with Integrated Water-Gas-Shift Membrane Reactor water-gas-shift.pdf (597.03 KB) More Documents & Publications ITP Energy Intensive Processes: Energy-Intensive Processes Portfolio: Addressing Key Energy Challenges Across U.S. Industry Energy-Intensive Processes Portfolio: Addressing Key Energy Challenges Across U.S. Industry CX-014220: Categorical Exclusion Determination

  10. Advanced dry head-end reprocessing of light water reactor spent...

    Office of Scientific and Technical Information (OSTI)

    reprocessing of light water reactor spent nuclear fuel Citation Details In-Document Search Title: Advanced dry head-end reprocessing of light water reactor spent nuclear fuel ...

  11. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  12. Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

    SciTech Connect

    Jason Hales; Various

    2014-06-01

    The US DOEs Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.

  13. Practical combinations of light-water reactors and fast reactors for future actinide transmutation

    SciTech Connect

    Collins, Emory D.; Renier, John-Paul

    2007-07-01

    Multicycle partitioning-transmutation (P-T) studies continue to show that use of existing light-water reactors (LWRs) and new advanced light-water reactors (ALWRs) can effectively transmute transuranic (TRU) actinides, enabling initiation of full actinide recycle much earlier than waiting for the development and deployment of sufficient fast reactor (FR) capacity. The combination of initial P-T cycles using LWRs/ALWRs in parallel with economic improvements to FR usage for electricity production, and a follow-on transition period in which FRs are deployed, is a practical approach to near-term closure of the nuclear fuel cycle with full actinide recycle. (authors)

  14. Improving Light Water Reactor Fuel Reliability Via Flow-Induced...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Improving Light Water Reactor Fuel Reliability Via Flow-Indu... Failures of the fuel rod elements used to power U.S. nuclear ... and a recognized bottleneck to optimal fuel utilization. ...

  15. Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles

    SciTech Connect

    Yin, Shengjun; Bass, Bennett Richard; Stevens, Gary; Kirk, Mark

    2011-01-01

    This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle3 3. PWR inlet nozzle; ; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; To assess the significance of attached piping loads on the stresses in the nozzle corner region; and To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.

  16. Process for treating effluent from a supercritical water oxidation reactor

    DOEpatents

    Barnes, C.M.; Shapiro, C.

    1997-11-25

    A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor. 6 figs.

  17. Process for treating effluent from a supercritical water oxidation reactor

    DOEpatents

    Barnes, Charles M.; Shapiro, Carolyn

    1997-01-01

    A method for treating a gaseous effluent from a supercritical water oxidation reactor containing entrained solids is provided comprising the steps of expanding the gas/solids effluent from a first to a second lower pressure at a temperature at which no liquid condenses; separating the solids from the gas effluent; neutralizing the effluent to remove any acid gases; condensing the effluent; and retaining the purified effluent to the supercritical water oxidation reactor.

  18. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect

    Mardiansah, Deby; Takaki, Naoyuki

    2010-06-22

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  19. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    SciTech Connect

    Bahri, Che Nor Aniza Che Zainul Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  20. Effects of Water Radiolysis in Water Cooled Reactors, NERI Proposal No.99-0010. Technical progress report

    SciTech Connect

    Pimblott, S.M.

    2000-04-01

    OAK B188 Effects of Water Radiolysis in Water Cooled Reactors, NERI Proposal No.99-0010. Technical progress report

  1. Light Water Reactor Sustainability Accomplishments Report

    SciTech Connect

    McCarthy, Kathryn A.

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  2. Feasibility study on the thorium fueled boiling water breeder reactor

    SciTech Connect

    PetrusTakaki, N.

    2012-07-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  3. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    SciTech Connect

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  4. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges

    Office of Energy Efficiency and Renewable Energy (EERE)

    The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR...

  5. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect

    Austin, W.; Brinkley, D.

    2010-05-05

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these

  6. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    SciTech Connect

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

    This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

  7. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    VERA Requirements Document (VRD) - Revision 1 Stephen M. Hess AMA Focus Area 30 March 2012 CASL-U-2011-0074-002 VERA Technical Requirements by Component Consortium for Advanced Simulation of LWRs ii CASL-U-2011-0074-002 REVISION LOG Revision Date Affected Pages Revision Description 0 2/25/2011 All Original Version 1 3/31/2012 Revision 1 1A 4/19/2012 all Revision 1A (Mario) Document pages that are: Export Controlled _______________None__________________________________ IP/Proprietary/NDA

  8. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    IC Workflow Project: Final Report Andrew Godfrey Advanced Modeling Applications Oak Ridge National Laboratory March 30 th , 2012 godfreyat@ornl.gov CASL-U-2011-0236-002 IC Workflow Project Final Report Consortium for Advanced Simulation of LWRs i CASL-U-2011-0236-002 REVISION LOG Revision Date Affected Pages Revision Description 0 12/31/2011 All Original Draft Report 1 1/25/2012 All Minor Revision 2 3/30/2012 All Include data from Phase 2 Document pages that are: Export Controlled

  9. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Proposed Test Stand Selection Criteria Jess Gehin, ORNL Steve Hess, EPRI Zeses Karoutas, Westinghouse Rose Montgomery, TVA Advanced Modeling Applications January 23, 2013 September 14, 2012 CASL-U-2012-0146-000 Proposed Test Stand Selection Criteria Consortium for Advanced Simulation of LWRs ii CASL-U-2012-0146-000 REVISION LOG Revision Date Affected Pages Revision Description 0 9/28/12 All Original Report 1 1/23/2013 All Sections 3, 4, and 5 removed to create an Unlimited version. Document

  10. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL-U-2013-0146-000 VERA Release Plan Revision 1 Stephen M. Hess, EPRI Rose Montgomery, TVA AMA Focus Area 26 April 2013 VERA Technical Requirements by Component Consortium for Advanced Simulation of LWRs ii CASL-8 REVISION LOG Revision Date Affected Pages Revision Description 0 1/31/2013 N/A Original Version 1 3/30/2013 4 / 5 - 7 / 12 / 13 - 17 / 19 Revision to incorporate VERA Aciton Matrix for initial Test Stand release (June) and Alpha release (September) in 2013. Document pages that are:

  11. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    EPRI Test Stand Report Results and Feedback from the EPRI Test Stand Brenden Mervin, EPRI Martin Pytel, EPRI Dennis Hussey, EPRI Stephen Hess, EPRI 1 August 2014 CASL-U-2014-0121-000 CASL-U-2014-0121-000-a -a Please complete sections appropriate for this record. REVISION LOG Revision Date Affected Pages Revision Description 0 8/1/2014 All Original Report 1 12/17/2014 REVISION LOG Changed control entries from TBD to N/A to approve for public release Document pages that are: Export Controlled

  12. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    L2:AMA.P9.01 Update to Industry Test Stand Experience Using VERA Stephen M. Hess, EPRI Dennis Hussey, EPRI Brenden Mervin, EPRI Martin Pytel, EPRI Rose Montgomery, TVA Fausto Franceschini, Westinghouse Electric Company LLC Advanced Modeling Applications 30 September 2014 CASL-U-2014-0187-000 Consortium for Advanced Simulation of LWRs ii CASL- U-2014-0187-000 A. REVISION LOG Revision Date Affected Pages Revision Description 0 9/30/2014 All Original Report Document pages that are: Export

  13. CASL-U-2015-0123-000 Tom Downar

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    23-000 Tom Downar University of Michigan April 19, 2015 Discussion of Pin-Resolved Validation CASL-U-2015-0123-000 Discussion of Pin-Resolved Validation Tom Downar INERI Meeting April 19 th , 2015 CASL-U-2015-0123-000 MPACT Verification / Validation Plan The V&V plan includes two principal components: - Verification: * Unit testing (Brendan Kochunas / Dan Jabaay) * Regression testing (Ben Collins: L3RTM.PRT.P10.04 MPACT Regression Test Harness) * Solution verification (Wang/Martin/Collins):

  14. High-Temperature Water-Gas Shift Membrane Reactor Study

    SciTech Connect

    Ciocco, M.V.; Iyoha, O.; Enick, R.M.; Killmeyer, R.P.

    2007-06-01

    NETL’s Office of Research and Development is exploring the integration of membrane reactors into coal gasification plants as a way of increasing efficiency and reducing costs. Water-Gas Shift Reaction experiments were conducted in membrane reactors at conditions similar to those encountered at the outlet of a coal gasifier. The changes in reactant conversion and product selectivity due to the removal of hydrogen via the membrane reactor were quantified. Research was conducted to determine the influence of residence time and H2S on CO conversion in both Pd and Pd80wt%Cu membrane reactors. Effects of the hydrogen sulfide-to-hydrogen ratio on palladium and a palladium-copper alloy at high-temperature were also investigated. These results were compared to thermodynamic calculations for the stability of palladium sulfides.

  15. Self-Sustaining Thorium Boiling Water Reactors

    SciTech Connect

    Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra; Seifried, Jeffrey E.; Zhang, Guanheng; Varela, Christopher R.; Fratoni, Massimiliano; Vijic, Jasmina J.; Downar, Thomas; Hall, Andrew; Ward, Andrew; Jarrett, Michael; Wysocki, Aaron; Xu, Yunlin; Kazimi, Mujid; Shirvan, Koroush; Mieloszyk, Alexander; Todosow, Michael; Brown, Nicolas; Cheng, Lap

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  16. Development of 1000 MWe Advanced Boiling Water Reactor

    SciTech Connect

    Kazuo Hisajima; Ken Uchida; Keiji Matsumoto; Koichi Kondo; Shigeki Yokoyama; Takuya Miyagawa [Toshiba Corporation (Japan)

    2006-07-01

    1000 MWe Advanced Boiling Water Reactor has only two main steam lines and six reactor internal pumps, whereas 1350 MWe ABWR has four main steam lines and ten reactor internal pumps. In order to confirm how the differences affect hydrodynamic conditions in the dome and lower plenum of the reactor pressure vessel, fluid analyses have been performed. The results indicate that there is not substantial difference between 1000 MWe ABWR and 1350 MWe ABWR. The primary containment vessel of the ABWR consists of the drywell and suppression chamber. The suppression chamber stores water to suppress pressure increase in the primary containment vessel and to be used as the source of water for the emergency core cooling system following a loss-of-coolant accident. Because the reactor pressure vessel of 1000 MWe ABWR is smaller than that of 1350 MWe ABWR, there is room to reduce the size of the primary containment vessel. It has been confirmed feasible to reduce inner diameter of the primary containment vessel from 29 m of 1350 MWe ABWR to 26.5 m. From an economic viewpoint, a shorter outage that results in higher availability of the plant is preferable. In order to achieve 20-day outage that results in 97% of availability, improvement of the systems for removal of decay heat is introduced that enables to stop all the safety-related decay heat removal systems except at the beginning of an outage. (authors)

  17. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    SciTech Connect

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  18. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  19. Assessment of light water reactor accident management programs and experience

    SciTech Connect

    Hammersley, R.J.

    1992-03-01

    The objective of this report is to provide an assessment of the current light water reactor experience regarding accident management programs and associated technology developments. This assessment for light water reactor (LWR) designs is provided as a resource and reference for the development of accident management capabilities for the production reactors at the Savannah River Site. The specific objectives of this assessment are as follows: 1. Perform a review of the NRC, utility, and industry (NUMARC, EPRI) accident management programs and implementation experience. 2. Provide an assessment of the problems and opportunities in developing an accident management program in conjunction or following the Individual Plant Examination process. 3. Review current NRC, utility, and industry technological developments in the areas of computational tools, severe accident predictive tools, diagnostic aids, and severe accident training and simulation.

  20. Mechanical design of a light water breeder reactor

    DOEpatents

    Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

    1976-01-01

    In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

  1. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  2. METHOD OF OPERATING A HEAVY WATER MODERATED REACTOR

    DOEpatents

    Vernon, H.C.

    1962-08-14

    A method of removing fission products from the heavy water used in a slurry type nuclear reactor is described. According to the process the slurry is steam distilled with carbon tetrachloride so that at least a part of the heavy water and carbon tetrachloride are vaporized; the heavy water and carbon tetrachloride are separated; the carbon tetrachloride is returned to the steam distillation column at different points in the column to aid in depositing the slurry particles at the bottom of the column; and the heavy water portion of the condensate is purified. (AEC)

  3. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  4. Fatigue and environmentally assisted cracking in light water reactors

    SciTech Connect

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1992-03-01

    Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B steel were determined under small-amplitude cyclic loading in HP water with {approx}300 pbb dissolved oxygen. Radiation-induced segregation and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence also have been investigated. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain-rate tensile tests were conducted on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  5. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  6. Use of Thorium in Light Water Reactors (Journal Article) | SciTech...

    Office of Scientific and Technical Information (OSTI)

    Use of Thorium in Light Water Reactors Citation Details In-Document Search Title: Use of Thorium in Light Water Reactors Thorium-based fuels can be used to reduce concerns related ...

  7. DOE-NE Light Water Reactor Sustainability Program and EPRI Long...

    Energy Saver

    2-24562 Revision 4 DOE-NE Light Water Reactor Sustainability Program and EPRI Long Term ... INLEXT-12-24562 Revision 4 DOE-NE Light Water Reactor Sustainability Program and EPRI ...

  8. Fe CASL-U-2014-0014-002 VERA Common Input User Manual Scott Palmtag

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL-U-2014-0014-002 i Consortium for Advanced Simulation of LWRs Fe CASL-U-2014-0014-002 VERA Common Input User Manual Scott Palmtag Andrew Godfrey February 23, 2015 CASL-U-2014-0014-002 ii Consortium for Advanced Simulation of LWRs Oak Ridge National Laboratory in partnership with Electric Power Research Institute Idaho National Laboratory Los Alamos National Laboratory Massachusetts Institute of Technology North Carolina State University Sandia National Laboratories Tennessee Valley Authority

  9. Upper internals arrangement for a pressurized water reactor

    DOEpatents

    Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

    2013-07-09

    In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

  10. CASL-U-2015-0035-000 High Fidelity Modeling of Pellet-Clad Interaction

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    MC2015 - Joint International Conference on Mathematics and Computation ... CASL (Consortium for Advanced Simulation of Light ... federally sponsored research in accordance with ...

  11. Multi-Applications Small Light Water Reactor - NERI Final Report

    SciTech Connect

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  12. Accident Performance of Light Water Reactor Cladding Materials

    SciTech Connect

    Nelson, Andrew T.

    2012-07-24

    During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200 C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.

  13. Pebble Bed Boiling Water Reactor Concept With Superheated Steam

    SciTech Connect

    Tsiklauri, G.; Newman, D.; Meriwether, G.; Korolev, V. [Pacific Northwest National Laboratory, P.O. Box 999 Richland, WA 99352 (United States)

    2002-07-01

    An Advanced Nuclear Reactor concept is presented which extends Boiling Water Reactor technology with micro-fuel elements (MFE) and produces superheated steam. A nuclear plant with MFE is highly efficient and safe, due to ceramic-clad nuclear fuel. Water is used as both moderator and coolant. The fuel consists of spheres of about 1.5 mm diameter of UO{sub 2} with several external coatings of different carbonaceous materials. The outer coating of the particles is SiC, manufactured with chemical vapor disposition (CVD) technology. Endurance of the integrity of the SiC coating in water, air and steam has been demonstrated experimentally in Germany, Russia and Japan. This paper describes a result of a preliminary design and analysis of 3750 MWt (1500 MWe) plant with standard pressure of 16 MPa, which is widely achieved in the vessel of pressurized-water type reactors. The superheated steam outlet temperature of 550 deg. C elevates the steam cycle to high thermal efficiency of 42%. (authors)

  14. Analysis of scrams and forced outages at boiling water reactors

    SciTech Connect

    Earle, R. T.; Sullivan, W. P.; Miller, K. R.; Schwegman, W. J.

    1980-07-01

    This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability.

  15. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  16. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    1-000 MCNPX and KENO-VI Simulations of WBNP Unit 1 with Keno Comparisons Benchmark Results VERA Core Physics Progression Problems 1 - 5 William Gurecky Erich Schneider The University of Texas at Austin February 20, 2015 MCNPX and KENO-VI Benchmark Results Consortium for Advanced Simulation of LWRs ii CASL-U-2015-0221-000 REVISION LOG Revision Date Affected Pages Revision Description Document pages that are: Export Controlled _____________None___________________________ IP/Proprietary/NDA

  17. CASL-U-2015-0015-000 Modeling Integral Fuel

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    5-000 Modeling Integral Fuel Burnable Absorbers Using the Method of Characteristics Erik Daniel Walker University of Tennessee, Knoxville December 1, 2014 CASL-U-2015-0015-000 University of Tennessee, Knoxville Trace: Tennessee Research and Creative Exchange Masters Theses Graduate School 12-2014 Modeling Integral Fuel Burnable Absorbers Using the Method of Characteristics Erik Daniel Walker University of Tennessee - Knoxville, ewalk@vols.utk.edu This Thesis is brought to you for free and open

  18. CASL-U-2015-0113-000 RPI Milestone:

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    13-000 RPI Milestone: Development of a Mechanistic Subcooled Boiling Model for PWR Assemblies M.Z. Podowski Rensselaer Polytechnic Institute (RPI) August 31, 2014 CASL-U-2015-0113-000 August 31, 2014 RPI Milestone: Development of a mechanistic subcooled boiling model for PWR assemblies 1 TOP FOCUS AREA ACHIEVEMENTS * Top achievement 1 - The formulation of a mechanistic multidimensional model of vapor condensation in subcooled boiling. The new model allows to separately capture each: the

  19. CASL-U-2015-0245-000 Materials

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    5-000 Materials Processing and Optimization (MPO) / Fuels, Materials and Chemistry (FMC) Zsolt Rak and Donald W. Brenner North Carolina State University July 7, 2015 CASL-U-2015-0245-000 NC STATE UNIVERSITY Materials Processing and Optimization (MPO) First-principles modeling of CRUD formation porous structure of agglomerated particulates; primarily NiFe 2 O 4 and NiO METHOD: semi-empirical thermodynamics based on the Density Functional Theory (DFT) * chemical and magnetic disorder are modeled

  20. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    DOEpatents

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  1. Transpiring wall supercritical water oxidation reactor salt deposition studies

    SciTech Connect

    Haroldsen, B.L.; Mills, B.E.; Ariizumi, D.Y.; Brown, B.G.

    1996-09-01

    Sandia National Laboratories has teamed with Foster Wheeler Development Corp. and GenCorp, Aerojet to develop and evaluate a new supercritical water oxidation reactor design using a transpiring wall liner. In the design, pure water is injected through small pores in the liner wall to form a protective boundary layer that inhibits salt deposition and corrosion, effects that interfere with system performance. The concept was tested at Sandia on a laboratory-scale transpiring wall reactor that is a 1/4 scale model of a prototype plant being designed for the Army to destroy colored smoke and dye at Pine Bluff Arsenal in Arkansas. During the tests, a single-phase pressurized solution of sodium sulfate (Na{sub 2}SO{sub 4}) was heated to supercritical conditions, causing the salt to precipitate out as a fine solid. On-line diagnostics and post-test observation allowed us to characterize reactor performance at different flow and temperature conditions. Tests with and without the protective boundary layer demonstrated that wall transpiration provides significant protection against salt deposition. Confirmation tests were run with one of the dyes that will be processed in the Pine Bluff facility. The experimental techniques, results, and conclusions are discussed.

  2. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect

    Not Available

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  3. Transactions of the nineteenth water reactor safety information meeting

    SciTech Connect

    Weiss, A.J.

    1991-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.

  4. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOEpatents

    Lau, Louis K. S.

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

  5. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  6. Materials Inventory Database for the Light Water Reactor Sustainability Program

    SciTech Connect

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  7. Boiling-Water Reactor internals aging degradation study. Phase 1

    SciTech Connect

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

  8. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    SciTech Connect

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  9. Camera Inspection Arm for Boiling Water Reactors - 13330

    SciTech Connect

    Martin, Scott; Rood, Marc

    2013-07-01

    Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

  10. Supercritical Water Reactor Cycle for Medium Power Applications

    SciTech Connect

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  11. CASL - Lift Forces in Bubbly Flows

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Lift Forces in Bubbly Flows The dynamics of two-phase (gas/liquid) bubbly flows are complex: bubbles deform and disperse; large latent heats and heat capacity differentials influence local boiling; and relatively small changes in heated surface temperatures yield order of magnitude changes in boiling complexity. Because the local void volume has a direct feedback effect on reactor neutron flux and fuel rod power production, prediction of local boiling rates and bulk boiling effects in nuclear

  12. CASL - The Michigan Parallel Characteristics Transport Code

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    The Michigan Parallel Characteristics Transport Code Verification of MPACT: The Michigan Parallel Characteristics Transport Code Benjamin Collins, Brendan Kochunas, Daniel Jabbay, Thomas Downar, William Martin Department of Nuclear Engineering and Radiological Sciences University of Michigan Andrew Godfrey Oak Ridge National Laboroatory MPACT (Michigan PArallel Characteristics Transport Code) is a new reactor analysis tool being developed at the University of Michigan as an advanced pin-resolved

  13. Self-Sustaining Thorium Boiling Water Reactors (Technical Report...

    Office of Scientific and Technical Information (OSTI)

    of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; ... Language: English Subject: 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ...

  14. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    SciTech Connect

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  15. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    SciTech Connect

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  16. Light Water Reactor Sustainability Constellation Pilot Project FY12 Summary Report

    SciTech Connect

    R. Johansen

    2012-09-01

    Summary report for Light Water Reactor Sustainability (LWRS) activities related to the R. E. Ginna and Nine Mile Point Unit 1 for FY12.

  17. Light Water Reactor Sustainability Constellation Pilot Project FY13 Summary Report

    SciTech Connect

    R. Johansen

    2013-09-01

    Summary report for Light Water Reactor Sustainability (LWRS) activities related to the R. E. Ginna and Nine Mile Point Unit 1 for FY13.

  18. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    U-2014-0114-000 ORIGEN Integration into MPACT Revision 0 July 4, 2014 Brendan Kochunas William A. Wieselquist Ang Zhu L3:RTM.SUP.P9.03 - ORIGEN Integration into MPACT Consortium for Advanced Simulation of LWRs ii CASL-U-2014-0114-000 REVISION LOG Revision Date Affected Pages Revision Description 0 7/4/2014 All Initial version Document pages that are: Export Controlled None IP/Proprietary/NDA Controlled None Sensitive Controlled None Requested Distribution: To: N/A Copy: N/A L3:RTM.SUP.P9.03 -

  19. CASL Plan of Record 2 (1/11-6/11)

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    8-2014-0231-001 VERA Problem 10: Restart and Shuffling in MPACT Revision 1 February 26, 2015 Brendan Kochunas Daniel Jabaay Thomas Downar CASL-U-2014-0231-001 L3: RTM.PRT.P9.04 - VERA Problem 10: Restart and Shuffling in MPACT Consortium for Advanced Simulation of LWRs ii REVISION LOG Revision Date Affected Pages Revision Description 0 12/4/2014 All Initial version 1 2/27/2015 All Fixed typos from copy-paste in Appendix A, added section 3.3, updated results in Section 5. Additions to Section

  20. CASL-U-2015-0151-000 SMR Fuel Cycle

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    1-000 SMR Fuel Cycle Optimization and Control Rod Depletion Using NESTLE and LWROPT Keith E. Ottinger, P. Eric Collins, Nicholas P. Luciano, and G. Ivan Maldonado University of Tennessee - Knoxville March 29, 2015 CASL-U-2015-0151-000 Advances in Nuclear Fuel Management V (ANFM 2015) Hilton Head Island, South Carolina, USA, March 29 - April 1, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) SMR FUEL CYCLE OPTIMIZATION AND CONTROL ROD DEPLETION USING NESTLE AND LWROPT Keith E.

  1. CASL-U-2015-0154-000 I

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    4-000 I 2 s-LWR Fuel Management Option for an 18-Month Cycle Length D. Salazar, F. Franceschini, P. Ferroni Westinghouse Electric Company LLC B. Petrovic Georgia Institute of Technology March 29, 2015 CASL-U-2015-0154-000 Advances in Nuclear Fuel Management V (ANFM 2015) Hilton Head Island, South Carolina, USA, March 29 - April 1, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) ©2015 Westinghouse Electric Company LLC. All Rights Reserved, p. 1/11 I 2 S-LWR FUEL MANAGEMENT

  2. CASL-U-2015-0157-000 Stabilization Methods

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    7-000 Stabilization Methods for CMFD Acceleration M. Jarrett, B. Kelley, B. Kochunas, T. Downar, E. Larsen University of Michigan April 19, 2015 CASL-U-2015-0157-000 ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, TN * April 19-23, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) STABILIZATION METHODS FOR CMFD ACCELERATION M. Jarrett, B. Kelley, B.

  3. CASL-U-2015-0296-000 Document Progress Made

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL-U-2015-0296-000 Document Progress Made on Developing Multi- Phase Capabilities in Hydra-TH (LA-UR-15-26503, version 2) Balasubramanya T. Nadiga and Markus Berndt Los Alamos National Laboratory Andrew Bauer Kitware August 18, 2013 Document Progress on Developing Multi-Phase Capabilities in Hydra-TH 1 Balasubramanya T. Nadiga Markus Berndt Los Alamos National Laboratory Andrew Bauer Kitware Inc. 1 LA-UR-15-26503 Disclaimer: Los Alamos National Laboratory, an affirmative action/equal

  4. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    SciTech Connect

    Pan, Paul Y

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  5. CASL Milestone L3.AMA.VDT.P6.03

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    VERA (Problem 6) CASL-U-2013-0150-001 i Consortium for Advanced Simulation of LWRs Fe L3:AMA.VDT.P6.03 Coupled Single Assembly Solution with VERA (Problem 6), Rev 1 Scott Palmtag Core Physics July 31, 2013 CASL-U-2013-0150-001 Milestone L3:AMA.VDT.P6.03 CASL-U-2013-0150-001 ii Consortium for Advanced Simulation of LWRs Oak Ridge National Laboratory in partnership with Electric Power Research Institute Idaho National Laboratory Los Alamos National Laboratory Massachusetts Institute of Technology

  6. CASL Milestone L3.RTM.SUP.P8.02

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Implementation of the Transient Capability in MPACT: Phase II CASL-U-2014-0186-000 i Consortium for Advanced Simulation of LWRs Fe Implementation of the Transient Capability in MPACT: Phase II Yunlin Xu, Ang Zhu, Aaron Graham, Andrew Gerlach, Liangzhi Cao, Tom Downar, John Lee University of Michigan September 30, 2014 CASL-U-2011-0186-000 Milestone L4:RTM.SUP.P9.03 CASL-U-2014-0186-000 ii Consortium for Advanced Simulation of LWRs Oak Ridge National Laboratory in partnership with Electric Power

  7. CASL-8-2015-0103-000 Multi-Phase Flow: Direct Numerical Simulation

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    3-000 Multi-Phase Flow: Direct Numerical Simulation Igor Bolotnov North Carolina State University Gretar Tryggvason University of Notre Dame July 8-10, 2013 CASL-U-2015-0103-000 Multi-Phase Flow: Direct Numerical Simulation Multi-Phase Flow: Direct Numerical Simulation Igor Bolotnov - North Carolina State University Gretar Tryggvason - University of Notre Dame CASL Education Workshop, Oak Ridge National Laboratories, July 9-10, 2013 CASL-U-2015-0103-000 Multi-Phase Flow: Direct Numerical

  8. CASL - CFD-Based Turbulence Force Evaluation for Grid-to-Rod Fretting

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Phenomena CFD-Based Turbulence Force Evaluation for Grid-to-Rod Fretting Phenomena Article Background: Roger Lu - #83 - L1:CASL.P5.02 - Determine extent structural analysis amplifies (or damps) differences in pressure forces between different CFD codes for analysis of GTRF phenomenon (CASL.012) based on "CASL report CFD Turbulence Force Calculations and Grid-to-Rod Fretting Simulation" CFD-Based Turbulence Force Evaluation for Grid-to-Rod Fretting Phenomena R. Y. Lu and Z. Karoutas

  9. Revised accident source terms for light-water reactors

    SciTech Connect

    Soffer, L.

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  10. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    McCarthy, Kathryn A.; Busby, Jeremy; Hallbert, Bruce; Bragg-Sitton, Shannon; Smith, Curtis; Barnard, Cathy

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  11. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  12. Light Water Reactor Sustainability Program Integrated Program Plan

    SciTech Connect

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  13. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    SciTech Connect

    Hellesen, C.; Grape, S.; Haakanson, A.; Jacobson Svaerd, S.; Jansson, P.

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  14. Aging study of boiling water reactor high pressure injection systems

    SciTech Connect

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200{degrees}C (2,200{degrees}F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission`s Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed.

  15. Technologies for Upgrading Light Water Reactor Outlet Temperature

    SciTech Connect

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

  16. Pressurized-water reactor internals aging degradation study. Phase 1

    SciTech Connect

    Luk, K.H.

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  17. #LabChat: Supercomputing Our Way to the Future, Sept. 19 at 1...

    Office of Environmental Management (EM)

    nuclear reactors. Jaguar powers the virtual reactor at the Consortium for Advanced Simulation of Light Water Reactors (CASL). | Photo courtesy of Oak Ridge National Lab. ...

  18. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    SciTech Connect

    Lewis, M.R.

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  19. Light-water breeder reactor (LWBR Development Program)

    DOEpatents

    Beaudoin, B.R.; Cohen, J.D.; Jones, D.H.; Marier, L.J. Jr.; Raab, H.F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  20. Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment

    Energy.gov [DOE]

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the...

  1. Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7 |

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Department of Energy Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7 Initial Modeling of a Pressurized Water Reactor Completed Using RELAP-7 January 29, 2013 - 12:06pm Addthis Schematic of the OECD PWR benchmark used in the initial RELAP-7 demonstration Schematic of the OECD PWR benchmark used in the initial RELAP-7 demonstration RELAP-7 is a nuclear reactor system safety analysis code. Development started in October 2011, and during the past quarter the initial

  2. SEIS for the Production of Tritium in a Commercial Light Water Reactor |

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration | (NNSA) SEIS for the Production of Tritium in a Commercial Light Water Reactor The NNSA, a semi-autonomous agency within DOE, has prepared a Final Supplemental Environmental Impact Statement (SEIS) to update the environmental analyses in DOE's 1999 EIS for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS; DOE/EIS-0288). The CLWR EIS addressed the production of tritium in Tennessee Valley Authority (TVA) reactors using

  3. CASL - Successful expanded prediction of the nature of CRUD found in

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    pressurized water reactor coolant Successful expanded prediction of the nature of CRUD found in pressurized water reactor coolant The deposition of CRUD (Chalk River Unidentified Deposits) on fuel rods and in other areas of light water reactor (LWR) coolant loops is a serious issue, with potential safety and power reduction implications. Its in-depth understanding is sorely needed as reactors move to higher power densities or change chemistry programs. Methodologies to date to predict CRUD

  4. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    SciTech Connect

    Corradini, M. L.

    2015-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  5. Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

    SciTech Connect

    Shwageraus, Eugene; Zhao Xianfeng; Driscoll, Michael J.; Hejzlar, Pavel; Kazimi, Mujid S.; Herring, J. Stephen

    2004-07-15

    A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with {sup 235}U is necessary, and the {sup 235}U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO{sub 2}-UO{sub 2}) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of {sup 233}U from the {sup 232}Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the {sup 233}U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible

  6. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    SciTech Connect

    Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  7. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    SciTech Connect

    Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  8. Advanced dry head-end reprocessing of light water reactor spent...

    Office of Scientific and Technical Information (OSTI)

    Patent: Advanced dry head-end reprocessing of light water reactor spent nuclear fuel Citation Details In-Document Search Title: Advanced dry head-end reprocessing of light water ...

  9. Recent performance experience with US light water reactor self-actuating safety and relief valves

    SciTech Connect

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  10. CASL - Special Issue of the JOM: The Member Journal of TMS

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Special Issue of the JOM: The Member Journal of TMS vol 63 (no.8), August 2011 Editors: Brian Wirth (UTK - ORNL), Chris Stanek (LANL) and Kurt Edsinger (EPRI) CASL-MPO team members served as guest editors for vol 63, issue 8 of JOM (Journal of TMS, the Minerals, Metal and Materials Society) dedicated to Advanced Fuel Performance: Modeling and Simulation. The authors of these articles represent the formation of an industry, university and national lab team under CASL. In addition to Perspective

  11. Hydrogen water chemistry for BWRs (boiling water reactors): Materials behavior: Interim report

    SciTech Connect

    Gordon, B.M.; Jewett, C.W.; Pickett, A.E.; Walker, W.L.; Indig, M.E.; Andresen, P.L.; Niedrach, L.W.; Davis, R.B.

    1987-03-01

    The objective of this research program is to provide test data to guide future actions by boiling water reactor (BWR) owners regarding the use of hydrogen additions to the feedwater to mitigate pipe cracking during power operation. Numerous laboratory testing methods and approaches are being utilized in this program to evaluate and quantify the effects of this hydrogen water chemistry (HWC) on the corrosion performance of reactor materials, including full-scale pipe testing, fatigue crack initiation and growth studies, constant load tests, electrochemical potential (ECP) measurements, constant extension rate technique (CERT) testing, straining electrode tests (SET), oxide film analysis, fracture mechanics studies, general corrosion investigations and bent beam tests. The results to date are summarized in this report and indicate that HWC (which implies an ECP of Type-304 stainless steel below -230 mV/sub SHE/ coupled with a low water conductivity) generally has a beneficial effect on the corrosion performance of BWR structural materials. Specifically, HWC mitigates intergranular stress corrosion cracking (IGSCC) initiation and propagation in piping, provides an improved margin against environmental cracking in carbon steel and low alloy steel, and does not promote environmental cracking in other materials. A measurable, but acceptable, increase in the initial general corrosion kinetics of carbon and low alloy steel also accompanies the use of HWC. The laboratory data, together with the in-reactor test results, clearly indicate that HWC is an effective method of reducing the likelihood and rate of BWR pipe cracking.

  12. Selection of a suitable reactor type for water desalination and power generation in Saudi Arabia

    SciTech Connect

    Hussein, F.M.

    1988-03-01

    Selection of a reactor type suitable for water desalination and power generation is a complex process that involves the evaluation of many criteria and requires the professional judgment of many experts in different fields. A reactor type that is suitable for one country might not be suitable for another. This is especially true in the case of Saudi Arabia because of its strategic location, the nature of its land and people, and its moderate technological situation. A detailed study using a computer code based on Saaty's mathematical pairwise comparison technique and developed in a previous study was carried out to find the most suitable reactor for water desalination and power generation in Saudi Arabia from among five potential types: boiling water reactors (BWRs), pressurized water reactors, CANDU heavy water reactors (HWRs), steam-generating heavy water reactors (SGHWRs), and high-temperature gas-cooled reactors. It was concluded that the CANDU HWR is the most suitable type for this purpose followed first by the BWR, then the SGHWR.

  13. Passive decay heat removal system for water-cooled nuclear reactors

    DOEpatents

    Forsberg, Charles W.

    1991-01-01

    A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

  14. Multi-Application Small Light Water Reactor Final Report

    SciTech Connect

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as

  15. Multi-cycle boiling water reactor fuel cycle optimization

    SciTech Connect

    Ottinger, K.; Maldonado, G.I.

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  16. SMR Deliverable Final

    Office of Environmental Management (EM)

    Department of Energy Secretary Moniz Visits Oak Ridge National Laboratory SLIDESHOW: Secretary Moniz Visits Oak Ridge National Laboratory Addthis Energy Secretary Moniz at CASL 1 of 25 Energy Secretary Moniz at CASL Secretary Moniz tours the Consortium for Advanced Simulation of Light Water Reactors (CASL) facility at Oak Ridge National Laboratory (ORNL). CASL is one of the Energy Department's Energy Innovation Hubs. Date taken: 2013-06-03 12:04 Energy Secretary Moniz at CASL 2 of 25 Energy

  17. PowerPoint Presentation

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Energy Innovation Hub: Delivering Multi-scale Multiphysics Solutions for Commercial Nuclear Industry Challenges Jess C. Gehin, CASL Director Brian Kendrick, CASL FMC Focus Area Kevin Clarno, CASL PHI Focus Area Jeffrey Secker, CASL CRUD CPI ANS Webinar October 21, 2015 2 What is the Consortium for the Advanced Simulation of Light Water Reactors? CASL is the first U.S. DOE Energy Innovation Hub connecting fundamental research and technology development through an integrated partnership of

  18. Nuclear Energy Institute

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    (NEI) Summit Presentation University-Industry- Laboratory Partnerships: Gauging Effectiveness Douglas Kothe, CASL Director Oak Ridge National Laboratory February 26, 2014 CASL-U-2014-0355-000 CASL-U-2014-0355-000 University-Industry-Laboratory Partnerships Gauging Effectiveness CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub Douglas B. Kothe Oak Ridge National Laboratory Director, CASL 9 th Nuclear Energy R&D Summit Nuclear Energy Institute

  19. SLIDESHOW: Secretary Moniz Visits Oak Ridge National Laboratory |

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Department of Energy Secretary Moniz Visits Oak Ridge National Laboratory SLIDESHOW: Secretary Moniz Visits Oak Ridge National Laboratory Addthis Energy Secretary Moniz at CASL 1 of 25 Energy Secretary Moniz at CASL Secretary Moniz tours the Consortium for Advanced Simulation of Light Water Reactors (CASL) facility at Oak Ridge National Laboratory (ORNL). CASL is one of the Energy Department's Energy Innovation Hubs. Date taken: 2013-06-03 12:04 Energy Secretary Moniz at CASL 2 of 25 Energy

  20. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    SciTech Connect

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  1. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    SciTech Connect

    Boing, L.E.; Henley, D.R. ); Manion, W.J.; Gordon, J.W. )

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  2. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  3. Radiological Control of Water in Reactor Pond of MR Reactor in NRC 'Kurchatov Institute', During Dismantling Work - 13462

    SciTech Connect

    Stepanov, Alexey; Simirsky, Yury; Semin, Ilya; Volkovich, Anatoly; Ivanov, Oleg

    2013-07-01

    The analysis of the activity and radionuclide composition of water from the MR reactor pond for α,β,γ-ray radionuclides was made. To solve this problem we use a wide range of laboratory equipment: gamma spectrometric complex, beta spectrometric complex, vacuum alpha spectrometer, and spectrometric complex with liquid scintillator. The water from MR reactor pond contains: Cs-137 (2,6*10{sup 2} Bq/g), Co-60(1,8 Bq/g), Sr-90 (1,0*10{sup 2} Bq/g), H-3 (7,0*10{sup 3} Bq/g), and components of nuclear fuel (U-232,U-234,U-235,U-236,U-238). Therefore the cleaning water from radioactivity waste occurs to be quite a complicated radiochemical task. (authors)

  4. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  5. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  6. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    SciTech Connect

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  7. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  8. EIS-0288: Production of Tritium in a Commercial Light Water Reactor

    Energy.gov [DOE]

    This Environmental Impact Statement for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS) evaluates the environmental impacts associated with producing tritium at one or more...

  9. DOE-NE Light Water Reactor Sustainability Program and EPRI Long...

    Office of Environmental Management (EM)

    DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program - Joint Research and Development Plan Nuclear power has contributed almost 20% of the total ...

  10. Development of Light Water Reactor Fuels with Enhanced Accident Tolerance – Report to Congress

    Energy.gov [DOE]

    This report provides DOE’s plan to develop light water reactor (LWR) fuels with enhanced accident tolerance in response to 2012 Congressional direction and funding authorization. The result of the...

  11. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  12. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    DOEpatents

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  13. Implementation Plan and Initial Development of Nuclear Concrete Materials Database for Light Water Reactor Sustainability Program

    Office of Energy Efficiency and Renewable Energy (EERE)

    The FY10 activities for development of a nuclear concrete materials database to support the Light Water Reactor Sustainability Program are summarized. The database will be designed and constructed...

  14. Light-water-reactor safety research program. Quarterly progress report, January-March 1980

    SciTech Connect

    Massey, W.E.; Kyger, J.A.

    1980-08-01

    This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1980 on water-reactor-safety problems. The research and development area covered is Transient Fuel Response and Fission-Product Release.

  15. Passive gamma analysis of the boiling-water-reactor assemblies

    DOE PAGES [OSTI]

    Vo, D.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S.; Trellue, H.; et al

    2016-06-04

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in themore » past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

  16. An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors

    SciTech Connect

    Menlove, Howard O; Lee, Sang - Yoon

    2009-01-01

    This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

  17. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    SciTech Connect

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  18. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    SciTech Connect

    Lin, Jerry Y.S.

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  19. Advanced Fuel Performance: Modeling and Simulation Light Water Reactor Fuel Performance:

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    63 No. 8 * JOM 49 www.tms.org/jom.html Advanced Fuel Performance: Modeling and Simulation Light Water Reactor Fuel Performance: Current Status, Challenges, and Future High Fidelity Modeling K. Edsinger, C.R. Stanek, and B.D. Wirth How would you... ...describe the overall signifcance of this paper? This paper provides a concise description of the nuclear fuel used in pressurized water nuclear reactors and the most commonly observed fuel failure mechanisms. ...describe this work to a materials

  20. Draft Supplemental Environmental Impact Statement for the Production of Tritium in a Commercial Light Water Reactor

    National Nuclear Security Administration (NNSA)

    FRONT COVER Draft Supplemental Environmental Impact Statement for the Production of Tritium in a Commercial Light Water Reactor U.S. Department of Energy National Nuclear Security Administration DOE/EIS-0288-S1 August 2014 ACRONYMS AND ABBREVIATIONS CFR Code of Federal Regulations CLWR commercial light water reactor CO2e carbon dioxide equivalent DOE U.S. Department of Energy EIS environmental impact statement EPA U.S. Environmental Protection Agency °F degrees Fahrenheit FR Federal Register

  1. Chapter 4: Advancing Clean Electric Power Technologies | Light Water Reactors Technology Assessment

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Light Water Reactors Chapter 4: Technology Assessments Past, Present, and Future of the Technology The world's first full-scale nuclear power plant (NPP) devoted exclusively to peacetime uses came online in 1957. Light water reactors (LWRs) are now a mature technology, with over 350 operational LWRs worldwide (Figure 4.M.1) and over 60 under construction (Figure 4.M.2). 1 Note that the Fukushima accident adversely affected nuclear power operations in Japan (and other countries throughout the

  2. Slide 1

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Toward Predictive Modeling of Nuclear Reactor Performance: Application Development Experiences, Challenges, and Plans in CASL Presented at Louisiana State University, Baton Rouge, Louisiana Doug Kothe Oak Ridge National Laboratory February 27, 2014 CASL-U-2014-0359-000 CASL-U-2014-0359-000 Toward Predictive Modeling of Nuclear Reactor Performance: Application Development Experiences, Challenges, and Plans in CASL The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy

  3. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  4. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    SciTech Connect

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  5. Microsoft Word - NURETH-15 paper_587_formatted

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    ... Advanced Simulations of Light Water Reactors (CASL). 6. REFERENCES 1 B. ... during Loss-of-Flow Accident in Gen-IV Sodium Fast Reactor, Nuclear Technology, Vol. 183, ...

  6. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    SciTech Connect

    Greene, S.R.; Spellman, D.J.; Bevard, B.B.

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  7. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    SciTech Connect

    Thomas, Kenneth; Oxstrand, Johanna

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  8. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A; Matlack, Katie; Ramuhalli, Pradeep; Light, Glenn

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  9. Neutronic Study of Slightly Modified Water Reactors and Application to Transition Scenarios

    SciTech Connect

    Chambon, Richard; Guillemin, Perrine; Nuttin, Alexis; Bidaud, A.

    2007-07-01

    In this paper we have studied slightly modified water reactors and their applications to transition scenarios. The PWR and CANDU reactors have been considered. New fuels based on Thorium have been tested: Thorium/Plutonium and Thorium/Uranium- 233, with different fissile isotope contents. Changes in the geometry of the assemblies were also explored to modify the moderation ratio, and consequently the neutron flux spectrum. A core equivalent assembly methodology was introduced as an exploratory approach and to reduce the computation time. Several basic safety analyses were also performed. We have finally developed a new scenario code, named OSCAR (Optimized Scenario Code for Advanced Reactors), to study the efficiency of these modified reactors in transition to Gen IV reactors or in symbiotic fleet. (authors)

  10. The Virtual Environment for Reactor Applications (VERA): Design and architecture

    DOE PAGES [OSTI]

    Turner, John A.; Clarno, Kevin; Sieger, Matt; Bartlett, Roscoe; Collins, Benjamin; Pawlowski, Roger; Schmidt, Rodney; Summers, Randall

    2016-09-08

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL), the first DOE Hub, which was established in July 2010 for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both amore » software and a numerical perspective, along with the goals and constraints that drove the major design decisions and their implications. As a result, we explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the application of VERA tools for a variety of challenging problems within the nuclear industry.« less

  11. Milestone L1:CASL.P7.04 RSICC Release Testing Results Summary

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    L1:CASL.P7.04 RSICC Release Testing Results Summary Andrew Godfrey Oak Ridge National Laboratory September 30, 2013 CASL-U-2014-0009-000 Date: 9/30/2013 To: Matt Sieger c: Jess Gehin From: Andrew Godfrey Subject: RSICC Release Testing Results Matt, I have completed a thorough technical review of the VERA capabilities which are being released to RSICC. This memo provides documentation of the readiness of VERA for external release for several specific capabilities. In summary, I have found that

  12. CASL-U-2015-0150-000 Pellet-Cladding Mechanical Interaction

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    50-000 Pellet-Cladding Mechanical Interaction Analyses Using VERA Mervin, B. T., Pytel, M. L., Hussey, D. F., and Hess, S. M. Electric Power Research Institute April 1, 2015 CASL-U-2015-0150-000 PELLET-CLADDING MECHANICAL INTERACTION ANALYSES USING VERA Mervin, B. T., Pytel, M. L., Hussey, D. F., and Hess, S. M. Electric Power Research Institute 3420 Hillview Avenue, Palo Alto, CA 94304 bmervin@epri.com; mpytel@epri.com; dhussey@epri.com; shess@epri.com ABSTRACT A CASL Test Stand was launched in

  13. CASL - Initial Validation and Benchmark Study of new 3D CRUD Model

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Initial Validation and Benchmark Study of new 3D CRUD Model A new 3D CRUD model, known as "MAMBA" (for "MPO Advanced Model for Boron Analysis"), is being developed by the Crud Group within the MPO focus area of CASL. The 3D MAMBA v2.0 computer code was released to CASL on Feb. 28, 2012 and is capable of being run in "stand-alone" mode or in coupled mode with a thermal hydraulics computational fluid dynamics model (e.g., STAR-CCM+) and/or a neutron transport

  14. CASL Milestone L3.RTM.PRT.P7.05

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    4-0051-000 i Consortium for Advanced Simulation of LWRs Fe CASL-U-2014-0051-000 Demonstration of Neutronics Coupled to Thermal-Hydraulics for a Full-Core Problem using COBRA-TF/MPACT April 1, 2014 Brendan Kochunas Dan Jabaay Benjamin Collins Thomas Downar University of Michigan Milestone L3:RTM.P7.05 CASL-U-2014-0051-000 ii Consortium for Advanced Simulation of LWRs Oak Ridge National Laboratory in partnership with Electric Power Research Institute Idaho National Laboratory Los Alamos National

  15. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect

    Osborne, A.G.; Smith, T.A.; Deinert, M.R.

    2013-07-01

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  16. The Application of Structural Materials Data From the BN-350 Fast Reactor to Life Extension of Light Water Reactors

    SciTech Connect

    Romanenko, O.G.; Kislitsin, S.B.; Maksimkin, O.P.; Shiganakov, Sh.B.; Chumakov, Ye.V.; Dumchev, I.V.

    2006-07-01

    This paper describes the results of investigations of 08Cr16Ni11Mo3 (AISI 316 steel analogue) austenitic stainless steel irradiated in BN-350 breeder reactor at irradiation conditions close to that for Light Water Reactor (LWR) Internals. The pores were found in 08Cr16Ni11Mo3 steel irradiated at temperature 280 deg. C up to rather low damage 1.3 dpa and with dose rate 3.9 x 10{sup -9} dpa/s. There were obtained dose rate dependencies of yield strength, ultimate strength and ductility for 08Cr16Ni11Mo3 steel irradiated up to 7-13 dpa at 302-311 deg. C. These dependencies show a decrease in both yield strength and ultimate strength when dose rate decreases. There was observed an apparent decrease in total elongation when dose rate decreases, which was presumably connected with the pores formation in the steel at low dose rates. (authors)

  17. IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

    SciTech Connect

    Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

    2004-10-03

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.

  18. Nuclear reactor with makeup water assist from residual heat removal system

    DOEpatents

    Corletti, M.M.; Schulz, T.L.

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  19. Nuclear reactor with makeup water assist from residual heat removal system

    DOEpatents

    Corletti, Michael M.; Schulz, Terry L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  20. Generic component failure data base for light water and liquid sodium reactor PRAs (probabilistic risk assessments)

    SciTech Connect

    Eide, S.A.; Chmielewski, S.V.; Swantz, T.D.

    1990-02-01

    A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs). The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates. Using this approach, most of the failure rates are based on actual plant data rather than existing estimates. 21 refs., 9 tabs.

  1. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detector

    SciTech Connect

    Hellfeld, D.; Dazeley, S.; Bernstein, A.; Marianno, C.

    2015-11-25

    The potential of elastic antineutrino-electron scattering (ν¯e + e → ν¯e + e) in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor. Background was estimated via independent simulations and by appropriately scaling published measurements from similar detectors. Many potential backgrounds were considered, including solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclide and water-borne radon decays, and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. The detector response was modeled using a GEANT4-based simulation package. The results indicate that with the use of low radioactivity PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. The directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of theoretical conditions that, if satisfied in practice, would enable nuclear reactor antineutrino directionality in a Gd-doped water Cherenkov detector approximately 10 km from a large power reactor.

  2. Evolutionary/advanced light water reactor data report

    SciTech Connect

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  3. Microsoft Word - CASL Safety Relevance v4.docx

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    are then arranged into a grid assembly to enhance coolant mixing and to restrict fuel vibration and movement. The number of assemblies in a reactor core depends on reactor...

  4. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    SciTech Connect

    Holbrook, Mark; Kinsey, Jim

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  5. Tritium recovery from tritiated water with a two-stage palladium membrane reactor

    SciTech Connect

    Birdsell, S.A.; Willms, R.S.

    1997-04-01

    A process to recover tritium from tritiated water has been successfully demonstrated at TSTA. The 2-stage palladium membrane reactor (PMR) is capable of recovering tritium from water without generating additional waste. This device can be used to recover tritium from the substantial amount of tritiated water that is expected to be generated in the International Thermonuclear Experimental Reactor both from torus exhaust and auxiliary operations. A large quantity of tritiated waste water exists world wide because the predominant method of cleaning up tritiated streams is to oxidize tritium to tritiated water. The latter can be collected with high efficiency for subsequent disposal. The PMR is a combined catalytic reactor/permeator. Cold (non-tritium) water processing experiments were run in preparation for the tritiated water processing tests. Tritium was recovered from a container of molecular sieve loaded with 2,050 g (2,550 std. L) of water and 4.5 g of tritium. During this experiment, 27% (694 std. L) of the water was processed resulting in recovery of 1.2 g of tritium. The maximum water processing rate for the PMR system used was determined to be 0.5 slpm. This correlates well with the maximum processing rate determined from the smaller PMR system on the cold test bench and has resulted in valuable scale-up and design information.

  6. Evolution of the core physics concept for the Canadian supercritical water reactor

    SciTech Connect

    Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M.

    2013-07-01

    The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.

  7. Neutron collar calibration for assay of LWR (light-water reactor) fuel assemblies

    SciTech Connect

    Menlove, H.O.; Pieper, J.E.

    1987-03-01

    The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the /sup 235/U content, and the /sup 238/U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities.

  8. Experimental Breeder Reactor-II Primary Tank System Wash Water...

    Office of Environmental Management (EM)

    Pre-Developmental INL EBR-II Wash Water Treatment Technologies (PBS ADSHQTD0100 (0003199)) EBR-II Wash Water Workshop - The majority of the sodium has been removed, remaining ...

  9. Thermalhydraulic calculation for boiling water reactor and its natural circulation component

    SciTech Connect

    Trianti, Nuri Nurjanah,; Su’ud, Zaki; Arif, Idam; Permana, Sidik

    2015-09-30

    Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density and inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.

  10. Component failures at pressurized water reactors. Final report

    SciTech Connect

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.

  11. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    SciTech Connect

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  12. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect

    Ishii, M.; Xu, Y.; Revankar, S.T.

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  13. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    SciTech Connect

    C. L. Atwood; V. N. Shah; W. J. Galyean

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  14. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect

    Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  15. Consortium for Advanced Simulation of Light-Water Reactors To...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power ... University of Michigan; and the Idaho, Los Alamos, and Sandia national laboratories. ...

  16. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    SciTech Connect

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  17. A Qualitative Assessment of Thorium-Based Fuels in Supercritical Pressure Water Cooled Reactors

    SciTech Connect

    Weaver, Kevan Dean; Mac Donald, Philip Elsworth

    2002-10-01

    The requirements for the next generation of reactors include better economics and safety, waste minimization (particularly of the long-lived isotopes), and better proliferation resistance (both intrinsic and extrinsic). A supercritical pressure water cooled reactor has been chosen as one of the lead contenders as a Generation IV reactor due to the high thermal efficiency and compact/simplified plant design. In addition, interest in the use of thorium-based fuels for Generation IV reactors has increased based on the abundance of thorium, and the minimization of transuranics in a neutron flux; as plutonium (and thus the minor actinides) is not a by-product in the thorium chain. In order to better understand the possibility of the combination of these concepts to meet the Generation IV goals, the qualitative burnup potential and discharge isotopics of thorium and uranium fuel were studied using pin cell analyses in a supercritical pressure water cooled reactor environment. Each of these fertile materials were used in both nitride and metallic form, with light water reactor grade plutonium and minor actinides added. While the uranium-based fuels achieved burnups that were 1.3 to 2.7 times greater than their thorium-based counterparts, the thorium-based fuels destroyed 2 to 7 times more of the plutonium and minor actinides. The fission-to-capture ratio is much higher in this reactor as compared to PWR’s and BWR’s due to the harder neutron spectrum, thus allowing more efficient destruction of the transuranic elements. However, while the uranium-based fuels do achieve a net depletion of plutonium and minor actinides, the breeding of these isotopes limits this depletion; especially as compared to the thorium-based fuels.

  18. Application of the Isotope Ratio Method to a Boiling Water Reactor

    SciTech Connect

    Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Meriwether, George H.; Mitchell, Mark R.; Reid, Bruce D.

    2010-08-11

    The isotope ratio method is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods. All reactor materials contain trace elemental impurities at parts per million levels, and the isotopes of these elements are transmuted by neutron irradiation in a predictable manner. While measuring the change in a particular isotopes concentration is possible, it is difficult to correlate to energy production because the initial concentration of that element may not be accurately known. However, if the ratio of two isotopes of the same element can be measured, the energy production can then be determined without knowing the absolute concentration of that impurity since the initial natural ratio is known. This is the fundamental principle underlying the isotope ratio method. Extremely sensitive mass-spectrometric methods are currently available that allow accurate measurements of the impurity isotope ratios in samples. Additionally, indicator elements with stable activation products have been identified so that their post-irradiation isotope ratios remain constant. This method has been successfully demonstrated on graphite-moderated reactors. Graphite reactors are particularly well-suited to such analyses since the graphite moderator is resident in the fueled region of the core for the entire period of operation. Applying this method to other reactor types is more difficult since the resident portions of the reactor available for sampling are either outside the fueled region of the core or structural components of individual fuel assemblies. The goal of this research is to show that the isotope ratio method can produce meaningful results for light water-moderated power reactors. In this work, we use the isotope ratio method to estimate the energy

  19. Nanostructure of Metallic Particles in Light Water Reactor Used Nuclear Fuel

    SciTech Connect

    Buck, Edgar C.; Mausolf, Edward J.; Mcnamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2015-03-11

    The extraordinary nano-structure of metallic particles in light water reactor fuels points to possible high reactivity through increased surface area and a high concentration of high energy defect sites. We have analyzed the metallic epsilon particles from a high burn-up fuel from a boiling water reactor using transmission electron microscopy and have observed a much finer nanostructure in these particles than has been reported previously. The individual round particles that varying in size between ~20 and ~50 nm appear to consist of individual crystallites on the order of 2-3 nm in diameter. It is likely that in-reactor irradiation induce displacement cascades results in the formation of the nano-structure. The composition of these metallic phases is variable yet the structure of the material is consistent with the hexagonal close packed structure of epsilon-ruthenium. These findings suggest that unusual catalytic behavior of these materials might be expected, particularly under accident conditions.

  20. COLLOQUIUM: CASL: Consortium for Advanced Simulation of Light...

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    for Reactor Applications (VERA), incorporates science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and...

  1. Environmentally assisted cracking in light water reactors. Semiannual report, April--September 1991: Volume 13

    SciTech Connect

    Kassner, T F; Ruther, W E; Chung, H M; Hicks, P D; Hins, A G; Park, J Y; Soppet, W K; Shack, W J

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with {approx} 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  2. Coupled full core neutron transport/CFD simulations of pressurized water reactors

    SciTech Connect

    Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T.; Brewster, R.; Baglietto, E.; Yan, J.

    2012-07-01

    Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)

  3. Experimental techniques to determine salt formation and deposition in supercritical water oxidation reactors

    SciTech Connect

    Chan, J.P.C.; LaJeunesse, C.A.; Rice, S.F.

    1994-08-01

    Supercritical Water Oxidation (SCWO) is an emerging technology for destroying aqueous organic waste. Feed material, containing organic waste at concentrations typically less than 10 wt % in water, is pressurized and heated to conditions above water`s critical point where the ability of water to dissolve hydrocarbons and other organic chemicals is greatly enhanced. An oxidizer, is then added to the feed. Given adequate residence time and reaction temperature, the SCWO process rapidly produces innocuous combustion products. Organic carbon and nitrogen in the feed emerge as CO{sub 2} and N{sub 2}; metals, heteroatoms, and halides appear in the effluent as inorganic salts and acids. The oxidation of organic material containing heteroatoms, such as sulfur or phosphorous, forms acid anions. In the presence of metal ions, salts are formed and precipitate out of the supercritical fluid. In a tubular configured reactor, these salts agglomerate, adhere to the reactor wall, and eventually interfere by causing a flow restriction in the reactor leading to an increase in pressure. This rapid precipitation is due to an extreme drop in salt solubility that occurs as the feed stream becomes supercritical. To design a system that can accommodate the formation of these salts, it is important to understand the deposition process quantitatively. A phenomenological model is developed in this paper to predict the time that reactor pressure begins to rise as a function of the fluid axial temperature profile and effective solubility curve. The experimental techniques used to generate effective solubility curves for one salt of interest, Na{sub 2}SO{sub 4}, are described, and data is generated for comparison. Good correlation between the model and experiment is shown. An operational technique is also discussed that allows the deposited salt to be redissolved in a single phase and removed from the affected portion of the reactor. This technique is demonstrated experimentally.

  4. Slide 1

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    -000 CASL: A DOE Energy Innovation Hub - RPI Lahey Seminar, Troy, New York Douglas B. Kothe, CASL Director Oak Ridge National Laboratory February 20, 2013 CASL-U-2015-0061-000 1 CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub Douglas B. Kothe Oak Ridge National Laboratory Director, CASL CASL-U-2015-0061-000 2 * A Different Approach - "Multi-disciplinary, highly collaborative teams ideally working under one roof to solve priority technology

  5. In-Reactor Oxidation of Zircaloy-4 Under Low Water Vapor Pressures

    SciTech Connect

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370C). Data from these tests will be used to support fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex- reactor test results were performed to evaluate the influence of irradiation.

  6. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    SciTech Connect

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  7. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    SciTech Connect

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  8. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    SciTech Connect

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  9. Structural integrity of water reactor pressure boundary components. Quarterly progress report Apr-Jun 80

    SciTech Connect

    Loss, F.J.

    1981-02-20

    This report describes progress in a continuing program to characterize material properties performance with respect to structural integrity of light water reactor pressure boundary components. Progress under fracture mechanics highlights J-R curve trends from low upper shelf A533-B weld deposits irradiated under the HSST program. Fatigue crack growth rates are being determined for a variety of pressure vessel and piping steels in simulated nuclear coolant environments. Three regions of crack growth behavior which have been associated with classical stress corrosion cracking and corrosion fatigue now have been clearly defined for reactor vessel steels. A theory of the influence of dissolved oxygen content in the fatigue crack growth in simulated PWR coolant is proposed. Work in radiation sensitivity describes recent progress in radiation studies involving reactor vessel steels in a coordinated IAEA program. Also reported is a notch ductility characterization of A508-2 forging steel with irradiation.

  10. Structural integrity of water reactor pressure boundary components. Quarterly progress report, April-June 1980

    SciTech Connect

    Loss, F.J.

    1981-02-20

    This report describes progress in a continuing program to characterize material properties performance with respect to structural integrity of light water reactor pressure boundary components. Progress under fracture mechanics highlights J-R curve trends from low upper shelf A533-B weld deposits irradiated under the HSST program. Fatigue crack growth rates are being determined for a variety of pressure vessel and piping steels in simulated nuclear coolant environments. Three regions of crack growth behavior which have been associated with classical stress corrosion cracking and corrosion fatigue now have been clearly defined for reactor vessel steels. A theory of the influence of dissolved oxygen content in the fatigue crack growth in simulated PWR coolant is proposed. Work in radiation sensitivity describes recent progress in radiation studies involving reactor vessel steels in a coordinated IAEA program. Also reported is a notch ductility characterization of A508-2 forging steel with irradiation.

  11. Environmentally assisted cracking in Light Water Reactors. Volume 16: Semiannual report, October 1992--March 1993

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  12. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail,

    2010-12-23

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this

  13. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    SciTech Connect

    Not Available

    1986-09-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised.

  14. CASL-U-2015-0179-000 Direct, On-the-Fly

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    9-000 Direct, On-the-Fly Calculation of Unresolved Resonance Region Cross Sections in Monte Carlo Simulations Jonathan A. Walsh, Benoit Forget, and Kord S. Smith Massachusetts Institute of Technology Brian C. Kiedrowski and Forrest B. Brown Los Alamos National Laboratory April 19, 2015 CASL-U-2015-0179-000 ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, Tennessee *

  15. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    SciTech Connect

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  16. Slide 1

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    CASL/WEC Industry Fellows Forum Presentation Fausto Franceschini Westinghouse Electric Company October 8, 2014 CASL-U-2013-0330-000 1 Westinghouse Non-Proprietary Class 3 © 2013 Westinghouse Electric Company LLC. All Rights Reserved. CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub Fausto Franceschini Fellow Engineer Westinghouse Electric Company LLC Research and Technology CASL-U-2013-0330-000 2 Westinghouse Non-Proprietary Class 3 Outline *

  17. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    DOE PAGES [OSTI]

    Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2015-05-17

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

  18. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    SciTech Connect

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  19. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    SciTech Connect

    Budd, W.A.

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  20. Structural integrity of water reactor pressure boundary components. Annual report for 1983. Volume 2

    SciTech Connect

    Loss, F.J.

    1984-09-01

    The objective of this research program is to characterize materials behavior in relation to structural safety and reliability of pressure boundary components for light water reactors. Specific objectives include developing an understanding of elastic-plastic fracture and environmentally-assisted crack propagation phenomena in terms of continuum mechanics, metallurgical variables, and neutron irradiation. Emphasis is placed on identifying metallurgical factors responsible for radiation embrittlement of steels and on developing procedures for embrittlement relief, including guidelines for radiation-resistant steels. The underlying goal is the interpretation of material properties performance to establish engineering criteria for structural reliability and long-term operation. Current work is organized into three major tasks: (1) fracture mechanics investigations, (2) environmentally-assisted crack growth in high temperature, primary reactor water and (3) radiation sensitivity and postirradiation properties recovery. Research progress in these tasks for 1983 is summarized.

  1. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    SciTech Connect

    Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2015-05-17

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.

  2. EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor (CLWR) Tritium Readiness Supplemental Environmental Impact Statement

    Energy.gov [DOE]

    This Supplemental EIS updates the environmental analyses in DOE’s 1999 EIS for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS). The CLWR EIS addressed the production of tritium in Tennessee Valley Authority reactors in Tennessee using tritium-producing burnable absorber rods.

  3. EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor Supplemental Environmental Impact Statement

    Energy.gov [DOE]

    This Supplemental EIS updates the environmental analyses in DOE’s 1999 EIS for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS). The CLWR EIS addressed the production of tritium in Tennessee Valley Authority reactors in Tennessee using tritium-producing burnable absorber rods.

  4. Nondestructive verification with minimal movement of irradiated light-water-reactor fuel assemblies

    SciTech Connect

    Phillips, J.R.; Bosler, G.E.; Halbig, J.K.; Klosterbuer, S.F.; Menlove, H.O.

    1982-10-01

    Nondestructive verification of irradiated light-water reactor fuel assemblies can be performed rapidly and precisely by measuring their gross gamma-ray and neutron signatures. A portable system measured fuel assemblies with exposures ranging from 18.4 to 40.6 GWd/tU and with cooling times ranging from 1575 to 2638 days. Differences in the measured results for side or corner measurements are discussed. 25 figures, 20 tables.

  5. Computational fluid dynamic analysis of a closure head penetration in a pressurized water reactor

    SciTech Connect

    Forsyth, D.R.; Schwirian, R.E.

    1995-09-01

    ALLOY 600 has been used typically for penetrations through the closure head in pressurized water reactors because of its thermal compatibility with carbon steel, superior resistance to chloride attack and higher strength than the austenitic stainless steels. Recent plant operating experience with this alloy has indicated that this material may be susceptible to degradation. One of the major parameters relating to degradation of the head penetrations are the operational temperatures and stress levels in the penetration.

  6. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  7. Boiling Water Reactor Fuel Cycle Optimization for Prevention of Channel-Blade Interference

    SciTech Connect

    Kropaczek, David J.; Karve, Atul A.; Oyarzun, Christian C.; Asgari, Mehdi

    2006-07-01

    A formal optimization method for eliminating the potential of Boiling Water Reactor channel-blade interference is presented within the context of fuel cycle design. The method is based on the use of threshold constraints on blade force as penalty terms within an objective function that are employed as part of a search algorithm. Results demonstrate the effectiveness of the constraint formulation in eliminating channel-blade interference as part of the design of the core loading and operational strategy. (authors)

  8. Fuel assembly for the production of tritium in light water reactors

    DOEpatents

    Cawley, William E.; Trapp, Turner J.

    1985-01-01

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  9. Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

    SciTech Connect

    Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

    2011-08-01

    Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

  10. National Library of Energy (BETA): the Department of Energy's...

    Office of Scientific and Technical Information (OSTI)

    Extremes Center for Materials Science of Nuclear Fuel Center for Molecular ... of Light Water Reactors (CASL) (Nuclear Modeling and Simulation Energy Innovation ...

  11. PowerPoint Presentation

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Rod Cusping Treatment in MPACT CASL is currently developing a new core simulator called MPACT to solve neutron transport problems for light-water nuclear reactors....

  12. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    SciTech Connect

    Shannon M. Bragg-Sitton

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  13. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    SciTech Connect

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K.

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  14. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    SciTech Connect

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment.

  15. Advanced fuel assembly characterization capabilities based on gamma tomography at the Halden boiling water reactor

    SciTech Connect

    Holcombe, S.; Eitrheim, K.; Svaerd, S. J.; Hallstadius, L.; Willman, C.

    2012-07-01

    Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Inst. for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala Univ.. (authors)

  16. Modeling of the performance of weapons MOX fuel in light water reactors

    SciTech Connect

    Alvis, J.; Bellanger, P.; Medvedev, P.G.; Peddicord, K.L.; Gellene, G.I.

    1999-05-01

    Both the Russian Federation and the US are pursing mixed uranium-plutonium oxide (MOX) fuel in light water reactors (LWRs) for the disposition of excess plutonium from disassembled nuclear warheads. Fuel performance models are used which describe the behavior of MOX fuel during irradiation under typical power reactor conditions. The objective of this project is to perform the analysis of the thermal, mechanical, and chemical behavior of weapons MOX fuel pins under LWR conditions. If fuel performance analysis indicates potential questions, it then becomes imperative to assess the fuel pin design and the proposed operating strategies to reduce the probability of clad failure and the associated release of radioactive fission products into the primary coolant system. Applying the updated code to anticipated fuel and reactor designs, which would be used for weapons MOX fuel in the US, and analyzing the performance of the WWER-100 fuel for Russian weapons plutonium disposition are addressed in this report. The COMETHE code was found to do an excellent job in predicting fuel central temperatures. Also, despite minor predicted differences in thermo-mechanical behavior of MOX and UO{sub 2} fuels, the preliminary estimate indicated that, during normal reactor operations, these deviations remained within limits foreseen by fuel pin design.

  17. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    SciTech Connect

    Todosow M.; Todosow M.; Raitses, G. Galperin, A.

    2009-07-12

    Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

  18. Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

    SciTech Connect

    Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

    2004-10-06

    The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal

  19. Draft Supplemental Environmental Impact Statement for the Production of Tritium in a Commercial Light Water Reactor, Summary

    National Nuclear Security Administration (NNSA)

    Draft Supplemental Environmental Impact Statement for the Production of Tritium in a Commercial Light Water Reactor Summary U.S. Department of Energy National Nuclear Security Administration DOE/EIS-0288-S1 August 2014 ACRONYMS AND ABBREVIATIONS CFR Code of Federal Regulations CLWR commercial light water reactor DOE U.S. Department of Energy EIS environmental impact statement EPA U.S. Environmental Protection Agency FR Federal Register NEPA National Environmental Policy Act of 1969 NNSA National

  20. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    SciTech Connect

    Orlov, Andrey; Degueldre, Claude; Kaufmann, Wilfried

    2013-01-15

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  1. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    SciTech Connect

    Wheeler, Timothy A.; Liao, Huafei

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  2. Development of Mechanistic Modeling Capabilities for Local Neutronically-Coupled Flow-Induced Instabilities in Advanced Water-Cooled Reactors

    SciTech Connect

    Michael Podowski

    2009-11-30

    The major research objectives of this project included the formulation of flow and heat transfer modeling framework for the analysis of flow-induced instabilities in advanced light water nuclear reactors such as boiling water reactors. General multifield model of two-phase flow, including the necessary closure laws. Development of neurton kinetics models compatible with the proposed models of heated channel dynamics. Formulation and encoding of complete coupled neutronics/thermal-hydraulics models for the analysis of spatially-dependent local core instabilities. Computer simulations aimed at testing and validating the new models of reactor dynamics.

  3. Membrane contactor/separator for an advanced ozone membrane reactor for treatment of recalcitrant organic pollutants in water

    SciTech Connect

    Chan, Wai Kit; Joueet, Justine; Heng, Samuel; Yeung, King Lun; Schrotter, Jean-Christophe

    2012-05-15

    An advanced ozone membrane reactor that synergistically combines membrane distributor for ozone gas, membrane contactor for pollutant adsorption and reaction, and membrane separator for clean water production is described. The membrane reactor represents an order of magnitude improvement over traditional semibatch reactor design and is capable of complete conversion of recalcitrant endocrine disrupting compounds (EDCs) in water at less than three minutes residence time. Coating the membrane contactor with alumina and hydrotalcite (Mg/Al=3) adsorbs and traps the organics in the reaction zone resulting in 30% increase of total organic carbon (TOC) removal. Large surface area coating that diffuses surface charges from adsorbed polar organic molecules is preferred as it reduces membrane polarization that is detrimental to separation. - Graphical abstract: Advanced ozone membrane reactor synergistically combines membrane distributor for ozone, membrane contactor for sorption and reaction and membrane separator for clean water production to achieve an order of magnitude enhancement in treatment performance compared to traditional ozone reactor. Highlights: Black-Right-Pointing-Pointer Novel reactor using membranes for ozone distributor, reaction contactor and water separator. Black-Right-Pointing-Pointer Designed to achieve an order of magnitude enhancement over traditional reactor. Black-Right-Pointing-Pointer Al{sub 2}O{sub 3} and hydrotalcite coatings capture and trap pollutants giving additional 30% TOC removal. Black-Right-Pointing-Pointer High surface area coating prevents polarization and improves membrane separation and life.

  4. Update to Risk-Informed Pressurized Water Reactor Vessel 10 to 20 Year Inspection Interval Extension

    SciTech Connect

    Palm, Nathan A.; Bishop, Bruce A.; Boggess, Cheryl L.

    2006-07-01

    The Pressurized Water Reactor Owners Group (formerly the Westinghouse Owners Group (WOG)) methodology for extending the inservice inspection interval for welds in pressurized water reactor (PWR) reactor pressure vessel (RPV) was introduced as ICONE12-49429. The paper presented a risk informed basis for extending the interval between inspections from the current interval of 10 years to 20 years. In the paper presented at ICONE-12, results of pilot studies on typical Westinghouse and Combustion Engineering Nuclear Steam Supply System (NSSS) designs of PWR vessels showed that the change in risk associated with the proposed inspection interval extension was within the guidelines specified in the United States Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174 for an acceptably small change in risk. Since the methodology was originally presented, the evaluation has been updated to incorporate the latest changes in the NRC Pressurized Thermal Shock (PTS) Risk Reevaluation Program and expanded to include the Babcock and Wilcox NSSS RPV design. The results of these evaluations demonstrate that the proposed RPV inspection interval extension remains a viable option for the industry. The updates to the methodology and input, pilot plant evaluations, results, process for demonstrating applicability of the pilot plant analysis to non-pilot lead plants and lessons learned from the evaluations performed are summarized in this paper. (authors)

  5. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    SciTech Connect

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  6. Dose rate estimates from irradiated light-water-reactor fuel assemblies in air

    SciTech Connect

    Lloyd, W.R.; Sheaffer, M.K.; Sutcliffe, W.G.

    1994-01-31

    It is generally considered that irradiated spent fuel is so radioactive (self-protecting) that it can only be moved and processed with specialized equipment and facilities. However, a small, possibly subnational, group acting in secret with no concern for the environment (other than the reduction of signatures) and willing to incur substantial but not lethal radiation doses, could obtain plutonium by stealing and processing irradiated spent fuel that has cooled for several years. In this paper, we estimate the dose rate at various distances and directions from typical pressurized-water reactor (PWR) and boiling-water reactor (BWR) spent-fuel assemblies as a function of cooling time. Our results show that the dose rate is reduced rapidly for the first ten years after exposure in the reactor, and that it is reduced by a factor of {approx}10 (from the one year dose rate) after 15 years. Even for fuel that has cooled for 15 years, a lethal dose (LD50) of 450 rem would be received at 1 m from the center of the fuel assembly after several minutes. However, moving from 1 to 5 m reduces the dose rate by over a factor of 10, and moving from 1 to 10 m reduces the dose rate by about a factor of 50. The dose rates 1 m from the top or bottom of the assembly are considerably less (about 10 and 22%, respectively) than 1 m from the center of the assembly, which is the direction of the maximum dose rate.

  7. Transactions of the Twenty-First Water Reactor Safety Information Meeting

    SciTech Connect

    Monteleone, S.

    1993-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

  8. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, Paul R.

    1994-01-01

    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  9. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    DOEpatents

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  10. Design and optimization of a back-flow limiter for the high performance light water reactor

    SciTech Connect

    Fischer, Kai; Laurien, Eckart; Claas, Andreas G.; Schulenberg, Thomas

    2007-07-01

    Design and Analysis of a back-flow limiter are presented, which is implemented as a safety device in the four inlet lines of the Reactor Pressure Vessel (RPV) of the High Performance Light Water Reactor (HPLWR). As a passive component, the back-flow limiter has no moving parts and belongs to the group of fluid diodes. It has low flow resistance for regular operation condition and a high flow resistance when the flow direction is reversed which is the case if a break of the feedwater line occurs. The increased flow resistance is due to a substantially increased swirl for reverse flow condition. The design is optimized employing 1D flow analyses in combination with 3D CFD analyses with respect to geometrical modifications, like the nozzle shape and swirler angles. (authors)

  11. Transactions of the twenty-fifth water reactor safety information meeting

    SciTech Connect

    Monteleone, S.

    1997-09-01

    This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session.

  12. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    SciTech Connect

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  13. Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

    SciTech Connect

    Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok; Kim, Hho-Jung

    2002-07-15

    The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.

  14. Microstructural characteristics of PWR [pressurized water reactor] spent fuel relative to its leaching behavior

    SciTech Connect

    Wilson, C.N.

    1986-01-01

    Microstructural, compositional and thermochemical properties of spent nuclear fuel are discussed relative to its potential performance as a high-level waste form under proposed Nevada Nuclear Waste Storage Investigations Project tuff repository conditions. Pressurized water reactor spent fuel specimens with various artificially induced cladding defects were leach tested in deionized water and in a reference tuff groundwater under ambient hot cell air and temperature conditions. Greater fractional actinide release was observed with bare fuel than with clad fuel leached through a cladding defect. Congruent actinide release and preferential release of cesium and technetium were observed in both water types. Selected summary radionuclide release data are presented and correlated to pre- and post-test microstructural characterization data.

  15. Light Water Reactor Sustainability Program Support and Modeling for the Boiling Water Reactor Station Black Out Case Study Using RELAP and RAVEN

    SciTech Connect

    Diego Mandelli; Curtis Smith; Thomas Riley; John Schroeder; Cristian Rabiti; Aldrea Alfonsi; Joe Nielsen; Dan Maljovec; Bie Wang; Valerio Pascucci

    2013-09-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated. In order to evaluate the impact of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the impact of power uprate on the safety of a boiled water reactor system. The case study considered is a loss of off-site power followed by the loss of diesel generators, i.e., a station black out (SBO) event. Analysis is performed by using a thermo-hydraulic code, i.e. RELAP-5, and a stochastic analysis tool currently under development at INL, i.e. RAVEN. Starting from the event tree models contained in SAPHIRE, we built the input file for RELAP-5 that models in great detail system dynamics under SBO conditions. We also interfaced RAVEN with RELAP-5 so that it would be possible to run multiple RELAP-5 simulation runs by changing specific keywords of the input file. We both employed classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. We also employed advanced data analysis and visualization tools that helped us to correlate simulation outcome such as maximum core temperature with a set of input uncertain parameters. Results obtained gave a detailed overview of the issues associated to power uprate for a SBO accident scenario. We were able to quantify how timing of safety related events were impacted by a higher reactor core power. Such insights can provide useful material to the decision makers to perform risk-infomed safety margins management.

  16. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    SciTech Connect

    Bhatti, Zaki; Hyland, B.; Edwards, G.W.R.

    2013-07-01

    The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction β) for coolant voiding as standard NU fuel. (authors)

  17. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    SciTech Connect

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs.

  18. Partial Defect Verification of the Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2010-02-05

    The International Atomic Energy Agency (IAEA) has the responsibility to carry out independent inspections of all nuclear material and facilities subject to safeguards agreements in order to verify compliance with non-proliferation commitments. New technologies have been continuously explored by the IAEA and Member States to improve the verification measures to account for declared inventory of nuclear material and detect clandestine diversion and production of nuclear materials. Even with these efforts, a technical safeguards challenge has remained for decades for the case of developing a method in identifying possible diversion of nuclear fuel pins from the Light Water Reactor (LWR) spent fuel assemblies. We had embarked on this challenging task and successfully developed a novel methodology in detecting partial removal of fuel from pressurized water reactor spent fuel assemblies. The methodology uses multiple tiny neutron and gamma detectors in the form of a cluster and a high precision driving system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly without any movement of the fuel. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. The combined information of gamma and neutron signature is used to produce base signatures and they are principally dependent on the geometry of the detector locations, and exhibit little sensitivity to initial enrichment, burn-up or cooling time. A small variation in the fuel bundle such as a few missing pins changes the shape of the signature to enable detection. This resulted in a breakthrough method which can be used to detect pin diversion without relying on the nuclear power plant operator's declared operation data. Presented are the results of various Monte Carlo simulation studies and experiments from actual commercial PWR spent fuel assemblies.

  19. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  20. Swelling in light water reactor internal components: Insights from computational modeling

    SciTech Connect

    Stoller, Roger E.; Barashev, Alexander V.; Golubov, Stanislav I.

    2015-08-01

    A modern cluster dynamics model has been used to investigate the materials and irradiation parameters that control microstructural evolution under the relatively low-temperature exposure conditions that are representative of the operating environment for in-core light water reactor components. The focus is on components fabricated from austenitic stainless steel. The model accounts for the synergistic interaction between radiation-produced vacancies and the helium that is produced by nuclear transmutation reactions. Cavity nucleation rates are shown to be relatively high in this temperature regime (275 to 325°C), but are sensitive to assumptions about the fine scale microstructure produced under low-temperature irradiation. The cavity nucleation rates observed run counter to the expectation that void swelling would not occur under these conditions. This expectation was based on previous research on void swelling in austenitic steels in fast reactors. This misleading impression arose primarily from an absence of relevant data. The results of the computational modeling are generally consistent with recent data obtained by examining ex-service components. However, it has been shown that the sensitivity of the model s predictions of low-temperature swelling behavior to assumptions about the primary damage source term and specification of the mean-field sink strengths is somewhat greater that that observed at higher temperatures. Further assessment of the mathematical model is underway to meet the long-term objective of this research, which is to provide a predictive model of void swelling at relevant lifetime exposures to support extended reactor operations.

  1. A study of out-of-phase power instabilities in boiling water reactors

    SciTech Connect

    March-Leuba, J.; Blakeman, E.D.

    1988-06-20

    This paper presents a study of the stability of subcritical neutronic modes in boiling water reactors that can result in out-of-phase power oscillations. A mechanism has been identified for this type of oscillation, and LAPUR code has been modified to account for it. Numerical results show that there is a region in the power-flow operating map where an out-or-phase stability mode is likely even if the core-wide mode is stable. 4 refs., 7 figs.

  2. Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

    SciTech Connect

    Le Pape, Yann; Huang, Hai

    2016-01-01

    Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

  3. CASL-U-2015-0132-000 Fausto Franceschini Westinghouse Electric Co. LLC,

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    32-000 Fausto Franceschini Westinghouse Electric Co. LLC, March 26, 2015 AP1000 ® PWR Startup Core Modeling and Simulation with VERA-CS CASL-U-2015-0132-000 F Advances in Nuclear Fuel Management V (ANFM 2015) Hilton Head Island, South Carolina, USA, March 29 - April 1, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) © 2015 Westinghouse Electric Company LLC. All Rights Reserved. p. 1/12 AP1000 ® PWR STARTUP CORE MODELING AND SIMULATION WITH VERA-CS F. Franceschini

  4. CASL-U-2015-0153-000 CRANE: A New Scale

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    3-000 CRANE: A New Scale Super-Sequence for Neutron Transport Calculations Congjian Wang and Hany S. Abdel-Khalik Purdue University Ugur Mertyurek Oak Ridge National Laboratory March 29, 2015 CASL-U-2015-0153-000 Advances in Nuclear Fuel Management V (ANFM 2015) Hilton Head Island, South Carolina, USA, March 29 - April 1, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015) © ANS 2015, Topical Meeting ANFM 2015, p. 1/11 CRANE: A NEW SCALE SUPER-SEQUENCE FOR NEUTRON TRANSPORT

  5. CASL-U-2015-0171-000 Improved Diffusion Coefficients for SPN Axial

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    1-000 Improved Diffusion Coefficients for SPN Axial Solvers In the MPACT 2D/1D Method Applied to the AP1000® PWR Start-Up Core Models Shane G. Stimpson, Benjamin Collins, Andrew Godfrey Oak Ridge National Laboratory Fausto Franceschini Westinghouse Electric Co., LLC Aaron Graham Thomas Downar University of Michigan April 19, 2015 CASL-U-2015-0171-000 ORNL is managed by UT-Battelle for the US Department of Energy Improved Diffusion Coefficients for SP N Axial Solvers In the MPACT 2D/1D Method

  6. CASL-U-2015-0178-000 An Azimuthal, Fourier Moment-based Transverse

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    8-000 An Azimuthal, Fourier Moment-based Transverse Leakage Approximation for the MPACT 2D/1D Method Shane Stimpson, Thomas Downar University of Michigan Benjamin Collins, Oak Ridge National Laboratory April 19, 2015 CASL-U-2015-0178-000 ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method * Nashville, TN * April 19-23, 2015, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2015)

  7. CASL - Effect of Grain Boundaries on Irradiation Growth of Zirconium-based

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Alloys Effect of Grain Boundaries on Irradiation Growth of Zirconium-based Alloys At the end of August, researchers S.I. Golubov, A.V. Barashev and R.E. Stoller delivered to CASL an analysis of the effect of grain size on the radiation growth of multigrain, hexagonal close-packed (hcp) metals, taking into account the features of cascade damage due to neutron exposure. Irradiation growth occurs in zirconium-based alloys used for LWR fuel cladding. Experimental data suggests that irradiation

  8. Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Kassner, T. F.; Ruther, W. E.; Shack, W. J.; Smith, J. L.; Soppet, W. K.; Strain; R. V.

    1999-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.

  9. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    SciTech Connect

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team indicates

  10. Examinations of Oxidation and Sulfidation of Grain Boundaries in Alloy 600 Exposed to Simulated Pressurized Water Reactor Primary Water

    SciTech Connect

    Schreiber, Daniel K.; Olszta, Matthew J.; Saxey, David W.; Kruska, Karen; Moore, K. L.; Lozano-Perez, Sergio; Bruemmer, Stephen M.

    2013-06-01

    High-resolution characterizations of intergranular attack in alloy 600 (Ni-17Cr-9Fe) exposed to 325 C simulated pressurized water reactor (PWR) primary water have been conducted using a combination of scanning electron microscopy, NanoSIMS, analytical transmission electron microscopy and atom probe tomography. The intergranular attack exhibited a two-stage microstructure that consisted of continuous corrosion/oxidation to a depth of ~200 nm from the surface followed by discrete Cr-rich sulfides to a further depth of ~500 nm. The continuous oxidation region contained primarily nanocrystalline MO-structure oxide particles and ended at Ni-rich, Cr-depleted grain boundaries with spaced CrS precipitates. Three-dimensional characterization of the sulfidized region using site-specific atom probe tomography revealed extraordinary grain boundary composition changes, including total depletion of Cr across a several nm wide dealloyed zone as a result of grain boundary migration.

  11. Environmentally assisted cracking in light water reactors - annual report, January-December 2001.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E; Hiller, R. W.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2003-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to {approx}2 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) ({approx}3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at {approx}325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600

  12. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOEpatents

    Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  13. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOEpatents

    Corletti, M.M.; Lau, L.K.; Schulz, T.L.

    1993-12-14

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

  14. Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    SciTech Connect

    Marshalkin, V. E. Povyshev, V. M.

    2015-12-15

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

  15. Evaluation of tubular reactor designs for supercritical water oxidation of U.S. Department of Energy mixed waste

    SciTech Connect

    Barnes, C.M.

    1994-12-01

    Supercritical water oxidation (SCWO) is an emerging technology for industrial waste treatment and is being developed for treatment of the US Department of Energy (DOE) mixed hazardous and radioactive wastes. In the SCWO process, wastes containing organic material are oxidized in the presence of water at conditions of temperature and pressure above the critical point of water, 374 C and 22.1 MPa. DOE mixed wastes consist of a broad spectrum of liquids, sludges, and solids containing a wide variety of organic components plus inorganic components including radionuclides. This report is a review and evaluation of tubular reactor designs for supercritical water oxidation of US Department of Energy mixed waste. Tubular reactors are evaluated against requirements for treatment of US Department of Energy mixed waste. Requirements that play major roles in the evaluation include achieving acceptable corrosion, deposition, and heat removal rates. A general evaluation is made of tubular reactors and specific reactors are discussed. Based on the evaluations, recommendations are made regarding continued development of supercritical water oxidation reactors for US Department of Energy mixed waste.

  16. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOEpatents

    Oosterkamp, Willem Jan; Marquino, Wayne

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

  17. Chimney for enhancing flow of coolant water in natural circulation boiling water reactor

    DOEpatents

    Oosterkamp, W.J.; Marquino, W.

    1999-01-05

    A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies is disclosed. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereas access to the fuel assemblies is not obstructed. 11 figs.

  18. Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

    SciTech Connect

    Mandelli, Diego; Prescott, Steven R; Smith, Curtis L; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J; Kinoshita, Robert A

    2011-07-01

    In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.

  19. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    SciTech Connect

    Not Available

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  20. FEASIBILITY OF RECYCLING PLUTONIUM AND MINOR ACTINIDES IN LIGHT WATER REACTORS USING HYDRIDE FUEL

    SciTech Connect

    Greenspan, Ehud; Todreas, Neil; Taiwo, Temitope

    2009-03-10

    The objective of this DOE NERI program sponsored project was to assess the feasibility of improving the plutonium (Pu) and minor actinide (MA) recycling capabilities of pressurized water reactors (PWRs) by using hydride instead of oxide fuels. There are four general parts to this assessment: 1) Identifying promising hydride fuel assembly designs for recycling Pu and MAs in PWRs 2) Performing a comprehensive systems analysis that compares the fuel cycle characteristics of Pu and MA recycling in PWRs using the promising hydride fuel assembly designs identified in Part 1 versus using oxide fuel assembly designs 3) Conducting a safety analysis to assess the likelihood of licensing hydride fuel assembly designs 4) Assessing the compatibility of hydride fuel with cladding materials and water under typical PWR operating conditions Hydride fuel was found to offer promising transmutation characteristics and is recommended for further examination as a possible preferred option for recycling plutonium in PWRs.

  1. Membrane reactor for water detritiation: a parametric study on operating parameters

    SciTech Connect

    Mascarade, J.; Liger, K.; Troulay, M.; Perrais, C.

    2015-03-15

    This paper presents the results of a parametric study done on a single stage finger-type packed-bed membrane reactor (PBMR) used for heavy water vapor de-deuteration. Parametric studies have been done on 3 operating parameters which are: the membrane temperature, the total feed flow rate and the feed composition through D{sub 2}O content variations. Thanks to mass spectrometer analysis of streams leaving the PBMR, speciation of deuterated species was achieved. Measurement of the amounts of each molecular component allowed the calculation of reaction quotient at the packed-bed outlet. While temperature variation mainly influences permeation efficiency, feed flow rate perturbation reveals dependence of conversion and permeation properties to contact time between catalyst and reacting mixture. The study shows that isotopic exchange reactions occurring on the catalyst particles surface are not thermodynamically balanced. Moreover, the variation of the heavy water content in the feed exhibits competition between permeation and conversion kinetics.

  2. Results from a scaled reactor cavity cooling system with water at steady state

    SciTech Connect

    Lisowski, D. D.; Albiston, S. M.; Tokuhiro, A.; Anderson, M. H.; Corradini, M. L.

    2012-07-01

    We present a summary of steady-state experiments performed with a scaled, water-cooled Reactor Cavity Cooling System (RCCS) at the Univ. of Wisconsin - Madison. The RCCS concept is used for passive decay heat removal in the Next Generation Nuclear Plant (NGNP) design and was based on open literature of the GA-MHTGR, HTR-10 and AVR reactor. The RCCS is a 1/4 scale model of the full scale prototype system, with a 7.6 m structure housing, a 5 m tall test section, and 1,200 liter water storage tank. Radiant heaters impose a heat flux onto a three riser tube test section, representing a 5 deg. radial sector of the actual 360 deg. RCCS design. The maximum heat flux and power levels are 25 kW/m{sup 2} and 42.5 kW, and can be configured for variable, axial, or radial power profiles to simulate prototypic conditions. Experimental results yielded measurements of local surface temperatures, internal water temperatures, volumetric flow rates, and pressure drop along the test section and into the water storage tank. The majority of the tests achieved a steady state condition while remaining single-phase. A selected number of experiments were allowed to reach saturation and subsequently two-phase flow. RELAP5 simulations with the experimental data have been refined during test facility development and separate effects validation of the experimental facility. This test series represents the completion of our steady-state testing, with future experiments investigating normal and off-normal accident scenarios with two-phase flow effects. The ultimate goal of the project is to combine experimental data from UW - Madison, UI, ANL, and Texas A and M, with system model simulations to ascertain the feasibility of the RCCS as a successful long-term heat removal system during accident scenarios for the NGNP. (authors)

  3. Comparative Study of Time-Domain versus Frequency-Domain Seismic Soil-Structure Interaction Analysis of Pressurized Water Reactor Containment Building

    Energy.gov [DOE]

    Comparative Study of Time-Domain versus Frequency-Domain Seismic Soil-Structure Interaction Analysis of Pressurized Water Reactor Containment Building

  4. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    SciTech Connect

    Rebak, Raul B.

    2014-12-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  5. Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

    SciTech Connect

    Thomas Baldwin; Magdy Tawfik; Leonard Bond

    2010-06-01

    In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R&D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10–12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I&C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy’s Light Water Reactor Sustainability Program. DOE has

  6. Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

    SciTech Connect

    Boucau, Joseph; Mirabella, C.; Nilsson, Lennart; Kreitman, Paul J.; Obert, Estelle

    2013-07-01

    Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French

  7. Apparatus for draining lower drywell pool water into suppresion pool in boiling water reactor

    DOEpatents

    Gluntz, Douglas M.

    1996-01-01

    An apparatus which mitigates temperature stratification in the suppression pool water caused by hot water drained into the suppression pool from the lower drywell pool. The outlet of a spillover hole formed in the inner bounding wall of the suppression pool is connected to and in flow communication with one end of piping. The inlet end of the piping is above the water level in the suppression pool. The piping is routed down the vertical downcomer duct and through a hole formed in the thin wall separating the downcomer duct from the suppression pool water. The piping discharge end preferably has an elevation at or near the bottom of the suppression pool and has a location in the horizontal plane which is removed from the point where the piping first emerges on the suppression pool side of the inner bounding wall of the suppression pool. This enables water at the surface of the lower drywell pool to flow into and be discharged at the bottom of the suppression pool.

  8. Slide 1

    U.S. Department of Energy (DOE) - all webpages (Extended Search)

    Guest Speaker Presentation at 69th Annual Meeting of the Oak Ridge Associated Universities (ORAU) Council of Sponsoring Institutions Doug Kothe Oak Ridge National Laboratory March 5-6, 2014 CASL-U-2014-0361-000 CASL-U-2014-0361-000 CASL: The Consortium for Advanced Simulation of Light Water Reactors A DOE Energy Innovation Hub Douglas B. Kothe Oak Ridge National Laboratory Director, CASL 69th Annual Meeting of the ORAU Council of Sponsoring Institutions March 5-6, 2014 CASL-U-2014-0361-000 2

  9. Annular seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen is a 35-element bundle made with a homogeneous mixture of reactor grade Pu and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several annular heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that the various core concepts can achieve a fissile utilization that is up to 30% higher than is currently achieved in a PT-HWR using conventional natural uranium fuel bundles. Up to 67% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 363 kg/year of U-233 is produced. Seed-blanket cores with ∼50% content of low-power blanket bundles may require power de-rating (∼58% to 65%) to avoid exceeding maximum limits for peak channel power, bundle power and linear element ratings. (authors)

  10. Checkerboard seed-blanket thorium fuel core concepts for heavy water moderated reactors

    SciTech Connect

    Bromley, B.P.; Hyland, B.

    2013-07-01

    New reactor concepts to implement thorium-based fuel cycles have been explored to achieve maximum resource utilization. Pressure tube heavy water reactors (PT-HWR) are highly advantageous for implementing the use of thorium-based fuels because of their high neutron economy and on-line re-fuelling capability. The use of heterogeneous seed-blanket core concepts in a PT-HWR where higher-fissile-content seed fuel bundles are physically separate from lower-fissile-content blanket bundles allows more flexibility and control in fuel management to maximize the fissile utilization and conversion of fertile fuel. The lattice concept chosen was a 35-element bundle made with a homogeneous mixture of reactor grade Pu (about 67 wt% fissile) and Th, and with a central zirconia rod to help reduce coolant void reactivity. Several checkerboard heterogeneous seed-blanket core concepts with plutonium-thorium-based fuels in a 700-MWe-class PT-HWR were analyzed, using a once-through thorium (OTT) cycle. Different combinations of seed and blanket fuel were tested to determine the impact on core-average burnup, fissile utilization, power distributions, and other performance parameters. It was found that various checkerboard core concepts can achieve a fissile utilization that is up to 26% higher than that achieved in a PT-HWR using more conventional natural uranium fuel bundles. Up to 60% of the Pu is consumed; up to 43% of the energy is produced from thorium, and up to 303 kg/year of Pa-233/U-233/U-235 are produced. Checkerboard cores with about 50% of low-power blanket bundles may require power de-rating (65% to 74%) to avoid exceeding maximum limits for channel and bundle powers and linear element ratings. (authors)

  11. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    SciTech Connect

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  12. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    SciTech Connect

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  13. Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed

    SciTech Connect

    Smirnov, A. Yu. Sulaberidze, G. A.; Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A. Proselkov, V. N.; Chibinyaev, A. V.

    2012-12-15

    A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

  14. Inhalation radiotoxicity of irradiated thorium as a heavy water reactor fuel

    SciTech Connect

    Edwards, G.W.R.; Priest, N.D.; Richardson, R.B.

    2013-07-01

    The online refueling capability of Heavy Water Reactors (HWRs), and their good neutron economy, allows a relatively high amount of neutron absorption in breeding materials to occur during normal fuel irradiation. This characteristic makes HWRs uniquely suited to the extraction of energy from thorium. In Canada, the toxicity and radiological protection methods dealing with personnel exposure to natural uranium (NU) spent fuel (SF) are well-established, but the corresponding methods for irradiated thorium fuel are not well known. This study uses software to compare the activity and toxicity of irradiated thorium fuel ('thorium SF') against those of NU. Thorium elements, contained in the inner eight elements of a heterogeneous high-burnup bundle having LEU (Low-enriched uranium) in the outer 35 elements, achieve a similar burnup to NU SF during its residence in a reactor, and the radiotoxicity due to fission products was found to be similar. However, due to the creation of such inhalation hazards as U-232 and Th-228, the radiotoxicity of thorium SF was almost double that of NU SF after sufficient time has passed for the decay of shorter-lived fission products. Current radio-protection methods for NU SF exposure are likely inadequate to estimate the internal dose to personnel to thorium SF, and an analysis of thorium in fecal samples is recommended to assess the internal dose from exposure to this fuel. (authors)

  15. Environmentally assisted cracking in light water reactors : semiannual report, July 2000 - December 2000.

    SciTech Connect

    Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.; Energy Technology

    2002-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on the mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to {approx}2.0 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288 C was found to decrease the fracture toughness of austenitic SSs. Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range

  16. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    SciTech Connect

    Soewono, C. N.; Takaki, N.

    2012-07-01

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  17. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    SciTech Connect

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

  18. Evaluation of a dilute chemical decontaminant for pressurized heavy water reactors

    SciTech Connect

    Velmurugan, S.; Narasimhan, S.V.; Mathur, P.K.; Venkateswarlu, K.S. )

    1991-12-01

    In this paper a dilute chemical decontamination formulation based on ethylene diamine tetraacetic acid, oxalic acid, and citric acid is evaluated for its efficacy in removing oxide layers in a pressurized heavy water reactor (PHWR). An ion exchange system that is specifically suited for fission product-dominated contamination in a PHWR is suggested for the reagent regeneration stage of the decontamination process. An attempt has been made to understand the redeposition behavior of various isotopes during the decontamination process. The polarographic method of identifying the species formed in the dissolution process is explained. Electrochemical measurements are employed to follow the course of oxide removal during the dissolution process. Scanning electron micrographs of metal coupons before and after the dissolution process exemplify the involvement of base metal in the formation of a ferrous oxalate layer. Material compatibility tests between the decontaminant and carbon steel, Monel-400, and Zircaloy-2 are reported.

  19. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    SciTech Connect

    Woo, H.H.; Lu, S.C.

    1981-09-15

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

  20. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    DOEpatents

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.