WorldWideScience

Sample records for advanced nuclear reactors

  1. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  2. Advanced nuclear reactor public opinion project

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  3. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  4. Advanced gas cooled nuclear reactor materials evaluation and development program

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  5. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  6. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    Energy Technology Data Exchange (ETDEWEB)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  7. Advanced Space Nuclear Reactors from Fiction to Reality

    Science.gov (United States)

    Popa-Simil, L.

    The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

  8. Advanced nuclear reactor public opinion project. Interim report

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  9. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Science.gov (United States)

    2010-01-01

    ... fuel and nuclear waste. 71.97 Section 71.97 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of transportation of nuclear waste was published in the Federal Register on June...

  10. Advanced methods for nuclear reactor gas laser coupling

    Energy Technology Data Exchange (ETDEWEB)

    Miley, G.H.; Verdeyen, J.T.

    1978-06-01

    Research is described that led to the discovery of three nuclear-pumped lasers (NPLs) using mixtures of Ne--N/sub 2/, He--Hg, and He or Ne with CO or CO/sub 2/. The Ne--N/sub 2/ NPL was the first laser obtained with modest neutron fluxes from a TRIGA reactor (vs fast burst reactors used elsewhere in such work), the He--Hg NPL was the first visible nuclear-pumped laser, while the Ne--CO and He--CO/sub 2/ lasers are the first to provide energy storage on a millisecond time scale. Important potential applications of NPLs include coupling and power transmission from remote power stations such as nuclear plants in satellites and neutron-feedback operation of inertial confinement fusion plants.

  11. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  12. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  13. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  14. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  15. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  16. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  17. Meeting nuclear data needs for advanced reactor system

    OpenAIRE

    Harada, H.; Shibata, K; Nishio, K.; IGASHIRA M.; PLOMPEN Arjan; Hambsch, Franz-Josef; Schillebeeckx, Peter; Gunsing, F.; Ledoux, X.; PALMIOTTI G.; Haight, R; ULLMANN J. L.; Tovesson, F.; Nelson, R.; Herman, M.

    2014-01-01

    The Working Party on International Nuclear Data Evaluation Co-operation (WPEC) has been established under the aegis of the OECD/NEA Nuclear Science Committee (NSC) to promote the exchange of information on nuclear data evaluations, validation and related topics. Its aim is also to provide a framework for co-operative activities between the members of the major nuclear data evaluation projects. This includes the possible exchange of scientists in order to encourage co-operation. Requirements f...

  18. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yang, Ying [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sridharan, K. [Univ. of Wisconsin, Madison, WI (United States)

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  19. Analysis of prospects for advanced nuclear reactors in western countries

    Energy Technology Data Exchange (ETDEWEB)

    Di Sapia, R. [ENEA, Rome (Italy). Area Energetica; Foskolos, K. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-05-01

    Nuclear energy deployment faces stagnation in western european and north american countries as a result of barriers that have appeared over the years. Such barriers were identified in the domains of economics, public acceptance, energy policy, technology, licensing and regulations as well as environment and waste disposal. It is to the nuclear community and particularly the industry to take the initiative and the leadership role for the most significant approaches to overcome these barriers. These approaches include concetration of efforts, lowering of costs and financial risks and extensive use of the experience accumulated so far; clear setting of priorities and long-term global consideration of the energy issue; encouraging an appropriate, stable regulatory environment and harmonisation of general safety objectives and principles, and adequate, globally consistent and clear information to the public. Also within the prime responsability of the nuclear community belong the safe operation of existing plants; making available all necessary information to the public, the media and the political leaders, supporting the development and execution of national energy polcies; supporting authorities in improving regulatory processes; taking all measures to improve economics of nuclear power; pursuing plans for the safe disposal of radioactive wastes. Governments should place energy issues in the appropriate priority level and encourage the establishment of an equally favourable environment for nuclear energy, including a greater consensus among controversial opinion representatives. Finally, authorities should established reasonable, transparent and predictable regulatory enviroments. This paper describes the barriers in a systematic way and proposes appropriate measures to overcame them.

  20. Nuclear design report for system-integrated modular advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Lee, Chung Chan; Zee, Sung Quun; Chang, Moon Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-02-01

    This report presents nuclear characteristics analysis results for SMART. Information is given on fuel loading, power density distributions, reactivity coefficients and control rod worths. The core consists of 57 modified Korean Standard Fuel Assemblies (m-KOFAs). and all fuel assemblies contain burnable absorbers to control the power distribution and the excess reactivity that is required for soluble boron-free and ultra longer cycle operation. The cycle length of SMART amounts to 990 EFPD corresponding to a cycle burnup of 26,160 MWD/MTU. 4 refs., 92 figs., 5 tabs. (Author)

  1. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  2. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    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

  3. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, September 23, 1976--December 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This report presents the results of work performed from September 23, 1976 through December 31, 1976 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  4. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-07

    The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  5. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-25

    Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

  6. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1979-June 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-25

    The results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  7. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-11-24

    Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. The activities associated with the characterization of the materials for the screening test program are reported, i.e., test specimen preparation, information from the materials characterization tests performed by General Electric, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented.

  8. Materials technology for an advanced space power nuclear reactor concept: Program summary

    Science.gov (United States)

    Gluyas, R. E.; Watson, G. K.

    1975-01-01

    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  9. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others

    2005-04-15

    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  10. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    Energy Technology Data Exchange (ETDEWEB)

    Marvin, M.D.

    1978-10-31

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated. (FS)

  11. Nuclear Reactor Engineering Analysis Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  12. Metallic fuels for advanced reactors

    Science.gov (United States)

    Carmack, W. J.; Porter, D. L.; Chang, Y. I.; Hayes, S. L.; Meyer, M. K.; Burkes, D. E.; Lee, C. B.; Mizuno, T.; Delage, F.; Somers, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

  13. Current status and future prospective of advanced radiation resistant oxide dispersion strengthened steel (ARROS) development for nuclear reactor system applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Kyu; Noh, Sang Hoon; Kang, Suk Hoon; Park, Jin Ju; Jin, Hyun Ju; Lee, Min Ku; Jang, Jin Sugn; Rhee, Chang Kyu [Nuclear Materials Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

  14. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  15. Advances in Nuclear Energy

    Science.gov (United States)

    Frois, B.

    2005-04-01

    This paper briefly reviews the next generations of nuclear reactors and the perspectives of development of nuclear energy. Advanced reactors will progressively replace the existing ones during the next two decades. Future systems of the fourth generation are planned to be built beyond 2030. These systems have been studied in the framework of the "Generation IV" International Forum. The goals of these systems is to have a considerable increase in safety, be economically competitive and produce a significantly reduced volume of nuclear wastes. The closed fuel cycle is preferred.

  16. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  17. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  18. Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

    Science.gov (United States)

    Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

    2013-01-01

    Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

  19. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  20. Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Niels Gronbech; Asta, Mark; Ozolins, Nigel Browning' Vidvuds; de Walle, Axel van; Wolverton, Christopher

    2011-12-29

    The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

  1. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  2. Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, K.; Clark, R.L.; Wood, R.T. [Oak Ridge National Lab., TN (United States)

    1994-04-01

    Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

  3. Measurement of fission cross-section of actinides at n_TOF for advanced nuclear reactors

    CERN Document Server

    Calviani, Marco; Montagnoli, G; Mastinu, P

    2009-01-01

    The subject of this thesis is the determination of high accuracy neutron-induced fission cross-sections of various isotopes - all of which radioactive - of interest for emerging nuclear technologies. The measurements had been performed at the CERN neutron time-of-flight facility n TOF. In particular, in this work, fission cross-sections on 233U, the main fissile isotope of the Th/U fuel cycle, and on the minor actinides 241Am, 243Am and 245Cm have been analyzed. Data on these isotopes are requested for the feasibility study of innovative nuclear systems (ADS and Generation IV reactors) currently being considered for energy production and radioactive waste transmutation. The measurements have been performed with a high performance Fast Ionization Chamber (FIC), in conjunction with an innovative data acquisition system based on Flash-ADCs. The first step in the analysis has been the reconstruction of the digitized signals, in order to extract the information required for the discrimination between fission fragm...

  4. Design of an Organic Simplified Nuclear Reactor

    OpenAIRE

    Koroush Shirvan; Eric Forrest

    2016-01-01

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attr...

  5. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  6. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  7. Design of an organic simplified nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

    2016-08-15

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  8. Design of an Organic Simplified Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-08-01

    Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  9. Improved best estimate plus uncertainty methodology including advanced validation concepts to license evolving nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Williams, Brian [Los Alamos National Laboratory; Mc Clure, Patrick [Los Alamos National Laboratory; Nelson, Ralph A [IDAHO NATIONAL LAB

    2010-01-01

    Many evolving nuclear energy programs plan to use advanced predictive multi-scale multi-physics simulation and modeling capabilities to reduce cost and time from design through licensing. Historically, the role of experiments was primary tool for design and understanding of nuclear system behavior while modeling and simulation played the subordinate role of supporting experiments. In the new era of multi-scale multi-physics computational based technology development, the experiments will still be needed but they will be performed at different scales to calibrate and validate models leading predictive simulations. Cost saving goals of programs will require us to minimize the required number of validation experiments. Utilization of more multi-scale multi-physics models introduces complexities in the validation of predictive tools. Traditional methodologies will have to be modified to address these arising issues. This paper lays out the basic aspects of a methodology that can be potentially used to address these new challenges in design and licensing of evolving nuclear technology programs. The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. An enhanced calibration concept is introduced and is accomplished through data assimilation. The goal is to enable best-estimate prediction of system behaviors in both normal and safety related environments. To achieve this goal requires the additional steps of estimating the domain of validation and quantification of uncertainties that allow for extension of results to areas of the validation domain that are not directly tested with experiments, which might include extension of the modeling and simulation (M&S) capabilities for application to full-scale systems. The new methodology suggests a formalism to quantify an adequate level of validation (predictive maturity) with respect to required selective data so that required testing can be minimized for cost

  10. Modeling Forced Flow Chemical Vapor Infiltration Fabrication of SiC-SiC Composites for Advanced Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Christian P. Deck

    2013-01-01

    Full Text Available Silicon carbide fiber/silicon carbide matrix (SiC-SiC composites exhibit remarkable material properties, including high temperature strength and stability under irradiation. These qualities have made SiC-SiC composites extremely desirable for use in advanced nuclear reactor concepts, where higher operating temperatures and longer lives require performance improvements over conventional metal alloys. However, fabrication efficiency advances need to be achieved. SiC composites are typically produced using chemical vapor infiltration (CVI, where gas phase precursors flow into the fiber preform and react to form a solid SiC matrix. Forced flow CVI utilizes a pressure gradient to more effectively transport reactants into the composite, reducing fabrication time. The fabrication parameters must be well understood to ensure that the resulting composite has a high density and good performance. To help optimize this process, a computer model was developed. This model simulates the transport of the SiC precursors, the deposition of SiC matrix on the fiber surfaces, and the effects of byproducts on the process. Critical process parameters, such as the temperature and reactant concentration, were simulated to identify infiltration conditions which maximize composite density while minimizing the fabrication time.

  11. Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

    Science.gov (United States)

    Mella, R.; Wenman, M. R.

    2013-06-01

    Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of

  12. IMPACT OF ENERGY GROUP STRUCTURE ON NUCLEAR DATA TARGET ACCURACY REQUIREMENTS FOR ADVANCED REACTOR SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    G. Palmiotti; M. Salvatores; H. Hiruta

    2011-06-01

    A target accuracy assessment study using both a fine and a broad energy structure has shown that less stringent nuclear data accuracy requirements are needed for the latter energy structure. However, even though a reduction is observed, still the requirements will be very difficult to be met unless integral experiments are also used to reduce nuclear data uncertainties. Target accuracy assessment is the inverse problem of the uncertainty evaluation. To establish priorities and target accuracies on data uncertainty reduction, a formal approach can be adopted by defining target accuracy on design parameters and finding out required accuracy on data in order to meet them. In fact, the unknown uncertainty data requirements can be obtained by solving a minimization problem where the sensitivity coefficients in conjunction with the constraints on the integral parameters provide the needed quantities for finding the solutions.

  13. Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

    2013-05-01

    The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

  14. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    Science.gov (United States)

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-01

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors. Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat. The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

  15. Comparison of Tungsten and Molybdenum Based Emitters for Advanced Thermionic Space Nuclear Reactors

    Science.gov (United States)

    Lee, Hsing H.; Dickinson, Jeffrey W.; Klein, Andrew C.; Lamp, Thomas R.

    1994-07-01

    Variations to the Advanced Thermionic Initiative thermionic fuel element are analyzed. Analysis included neutronic modeling with MCNP for criticality determination and thermal power distribution, and thermionic performance modeling with TFEHX. Changes to the original ATI configuration include the addition of W-HfC wire to the emitter for high temperature creep resistance improvement and substitution of molybdenum for the tungsten base material. Results from MCNP showed that all the tungsten used in the coating and base material must be 100% W-184 to obtain criticality. The presence of molybdenum in the emitter base affects the neutronic performance of the TFE by increasing the emitter neutron absorption cross section. Due to the reduced thermal conductivity for the molybdenum based emitter, a higher temperature is obtained resulting in a greater electrical power production. The thermal conductivity and resistivity of the composite emitter region were derived for the W-Mo composite and used in TFEHX.

  16. Teaching About Nature's Nuclear Reactors

    CERN Document Server

    Herndon, J M

    2005-01-01

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  17. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1--June 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-31

    The objectives of the program are to evaluate candidate alloys for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in the report includes the activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials receiver, and preliminary information from the materials characterization tests performed by General Electric. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The status of the data management system is also reviewed.

  18. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  19. Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  20. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  1. Helium mobility in advanced nuclear ceramics

    OpenAIRE

    Agarwal, Shradha

    2014-01-01

    While the current second and third generation nuclear plant designs provides an economically, technically, and publicly acceptable electricity supply in many markets, further advances in nuclear energy system design can broaden the opportunities for the use of nuclear energy. The fourth generation of nuclear reactors is under development. These new reactors are designed with the following objective in mind: sustainability, safety and reliability, economics, proliferation resistance. Out of si...

  2. Advances in nuclear science and technology

    CERN Document Server

    Greebler, Paul

    1968-01-01

    Advances in Nuclear Science and Technology Volume 4 provides information pertinent to the fundamental aspects of advanced reactor concepts. This book discusses the advances in various areas of general applicability, including modern perturbation theory, optimal control theory, and industrial application of ionizing radiations.Organized into seven chapters, this volume begins with an overview of the technology of sodium-cooled fast breeder power reactors and gas-cooled power reactors. This text then examines the key role of reactor safety in the development of fast breeder reactors. Other chapt

  3. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  4. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  5. A three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition in graphite components of advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, D.O.; Robinson, A.T.; Allen, D.A.; Picton, D.J.; Thornton, D.A. [TCS, Serco, Rutherford House, Olympus Park, Quedgeley, Gloucester, Gloucestershire GL2 4NF (United Kingdom); Shaw, S.E. [EDF Energy, Barnet Way, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2011-07-01

    This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition. (authors)

  6. Reactor antineutrinos and nuclear physics

    Science.gov (United States)

    Balantekin, A. B.

    2016-11-01

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states.

  7. Nuclear reactor downcomer flow deflector

    Energy Technology Data Exchange (ETDEWEB)

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  8. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  9. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-11-14

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.

  10. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1980-September 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-12

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, 950 and 1050/sup 0/C. Initiation of controlled purity helium creep-rupture testing in the intensive screening test program is discussed. In addition, the results of 1000-hour exposures at 750 and 850/sup 0/C on several experimental alloys are discussed.

  11. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  12. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  13. Uncertainty quantification approaches for advanced reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  14. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  15. Nuclear reactor effluent monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Minns, J.L.; Essig, T.H. [Nuclear Regulatory Commission, Washington, DC (United States)

    1993-12-31

    Radiological environmental monitoring and effluent monitoring at nuclear power plants is important both for normal operations, as well as in the event of an accident. During normal operations, environmental monitoring verifies the effectiveness of in-plant measures for controlling the release of radioactive materials in the plant. Following an accident, it would be an additional mechanism for estimating doses to members of the general public. This paper identifies the U.S. Nuclear Regulatory Commission (NRC) regulatory basis for requiring radiological environmental and effluent monitoring, licensee conditions for effluent and environmental monitoring, NRC independent oversight activities, and NRC`s program results.

  16. Advanced Non-Destructive Assessment Technology to Determine the Aging of Silicon Containing Materials for Generation IV Nuclear Reactors

    Science.gov (United States)

    Koenig, T. W.; Olson, D. L.; Mishra, B.; King, J. C.; Fletcher, J.; Gerstenberger, L.; Lawrence, S.; Martin, A.; Mejia, C.; Meyer, M. K.; Kennedy, R.; Hu, L.; Kohse, G.; Terry, J.

    2011-06-01

    To create an in-situ, real-time method of monitoring neutron damage within a nuclear reactor core, irradiated silicon carbide samples are examined to correlate measurable variations in the material properties with neutron fluence levels experienced by the silicon carbide (SiC) during the irradiation process. The reaction by which phosphorus doping via thermal neutrons occurs in the silicon carbide samples is known to increase electron carrier density. A number of techniques are used to probe the properties of the SiC, including ultrasonic and Hall coefficient measurements, as well as high frequency impedance analysis. Gamma spectroscopy is also used to examine residual radioactivity resulting from irradiation activation of elements in the samples. Hall coefficient measurements produce the expected trend of increasing carrier concentration with higher fluence levels, while high frequency impedance analysis shows an increase in sample impedance with increasing fluence.

  17. Conceptual Design of a Nuclear Reactor Dedicated for Desalination

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yong Hun; Moon, Jang Sik; Jeong, Yong Hoon [Korea Adavanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    The many advantages of nuclear desalination, the nuclear safety issues still remain a perennial problem today. To respond to such needs, the development of a desalination-dedicated nuclear reactor with maximized safety features was proposed. From the feasibility study, the desalination-dedicated reactor was found to be a good solution for meeting future water demand during the winter season in some countries like UAE by decoupling water and electricity supply. The economic analysis results indicated that under certain conditions, the desalination-dedicated reactor can produce freshwater at lower cost than the target nuclear cogeneration reactor using steam extraction technologies. A conceptual design of the desalination-dedicated nuclear reactor is in progress. The design features of the desalination-dedicated nuclear reactor could significantly enhance safety, reliability, and simplicity, and facilitate the extensive use of innovative passive safety systems. These maximized safety features of desalination-dedicated reactor could provide advanced capabilities for passive reactor shutdown and residual heat removal, and eventually prevent radioactivity release into the environment. The conceptual design achieved will provide a foothold for the future commercialization of the desalination-dedicated nuclear reactor and eventually help to address both a serious water crisis and nuclear safety issues.

  18. 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

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Liao, Huafei [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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.

  19. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  20. Development of CANDU advanced fuel fabrication technology - A development of amorphous alloys for the solder of nuclear reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jai Young; Lee, Ki Young; Kim, Yoon Kee; Jung, Jae Han; Yu, Ji Sang; Kim, Hae Yeol; Han, Young Su [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    In the case of advanced CANDU fuel being useful in future, the fabrication processes for soundness insurance of a improved nuclear fuel bundle must be developed at the same time because it have three times combustibility as existing fuel. In particular, as the improved nuclear fuel bundle in which a coated layer thickness is thinner than existing that, firmity of a joint part is very important. Therefore, we need to develop a joint technique using new solder which can settle a potential problem in current joining method. As the Zr-Be alloy system and the Ti-Be system are composed with the elements having high neutron permeability, they are suitable for joint of nuclear fuel pack. The various compositions Zr-Be and Ti-Be binary metallic glass alloys were applicable to the joining the nuclear fuel bundles. The thickness of joint layer using the Zr{sub 1-x} Be{sub x} amorphous ribbon as a solder is thinner than that using physical vapor deposited Be. Among the Zr{sub 1-x} Be{sub x} amorphous binary alloys, Zr{sub 0.7} Be{sub 0.3} binary alloy is the most appropriated for joint of nuclear fuel bundle because its joint layer is smooth and thin due to low degree of Be diffusion. The microstructures of brazed layer using Ti{sub 1-y} Be{sub y} alloy, however, a solid-solution layer composed with Zr and Ti is formed toward the Zr cladding sheath and many of Zr is detected in the joint lever. 20 refs., 8 tabs., 23 figs. (author)

  1. Reactors for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Angelo, J.A. Jr.

    1981-01-01

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

  2. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  3. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  4. Advanced Small Modular Reactor Economics Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J [ORNL

    2014-10-01

    This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic and nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation

  5. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

    Energy Technology Data Exchange (ETDEWEB)

    Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

    2005-07-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  6. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  7. Heat for industry from nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kikoin, I.K.; Novikov, V.M.

    Two factors which incline nations toward the use of heat from nuclear reactors for industrial use are: 1) exhaustion of cheap fossil fuel resources, and 2) ecological problems associated both with extraction of fossil fuel from the earth and with its combustion. In addition to the usual problems that beset nuclear reactors, special problems associated with using heat from nuclear reactors in various industries are explored.

  8. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation....

  9. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    Science.gov (United States)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  10. Radioactive target needs for nuclear reactor physics and nuclear astrophysics

    OpenAIRE

    Jurado, B.; Barreau, G.; Bacri, C. O.

    2010-01-01

    Nuclear Instruments and Methods in Physics Research Section A - In press.; Nuclear reaction cross sections of short-lived nuclei are key inputs for new generation nuclear reactor simulations and for models describing the nucleosynthesis of elements. After discussing various topics of nuclear astrophysics and reactor physics where the demand of nuclear data on unstable nuclei is strong, we describe the general characteristics of the targets needed to measure the requested data. In some cases t...

  11. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  12. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  13. New reactor technology: safety improvements in nuclear power systems.

    Science.gov (United States)

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  14. Fire Area Design of Advanced Pressurized Water Reactor Nuclear Plant%先进压水堆核电厂防火分区设计

    Institute of Scientific and Technical Information of China (English)

    陈闽烽; 张海波

    2014-01-01

    本文主要对先进压水堆核电厂防火分区设计的背景、目标、范围、原则及不同防火空间的特点进行了介绍,列举了重要厂房相关防火分区的设计情况。通过防火分区设计有助于火灾情况下保证安全相关设备的功能和人员的安全,提高了核电厂的总体防火安全水平。%This paper introduces the fire area design of advanced pressurized water reactor nuclear plant about background, objectives, scope, principles and characteristics of different fire area, and cites the fire area design of some important buildings. Through the design of fire area to help ensure the safety related equipment and personnel under fire and raise the overal level of fire safety in nuclear power plants.

  15. The Design of a Nuclear Reactor

    Indian Academy of Sciences (India)

    2016-09-01

    The aim of this largely pedagogical article is toemploy pre-college physics to arrive at an understanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuelpin, the moderator, and lastly the dimensions ofthe nuclear reactor.

  16. Advanced Burner Reactor Preliminary NEPA Data Study.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  17. On needs for particular isotopes with future advanced nuclear reactors; Quelques besoins en isotopes particuliers pour les reacteurs nucleaires avances du futur

    Energy Technology Data Exchange (ETDEWEB)

    Asou, M.; Porta, J. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Reacteurs

    1994-12-31

    This paper is concerned with two essential points with innovative approaches for nuclear reactors: utilization of {sup 15}N in uranium nitride fuels in order to avoid carbon 14 generation; utilization of rare earths, and especially gadolinium, for the control of the potential reactivity in the core in the case of a cycle lengthening from 12 to 18 months, and in the case of a water reactor operating without soluble boron. 1 tab., 17 refs.

  18. Structural materials challenges for advanced reactor systems

    Science.gov (United States)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  19. Nonlinear Ultrasonic Measurements in Nuclear Reactor Environments

    Science.gov (United States)

    Reinhardt, Brian T.

    Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this thesis, two ultrasonic characterization techniques will be explored. The first, finite amplitude wave propagation has been demonstrated to be sensitive to microstructural material property changes. It is a strong candidate to determine fuel evolution; however, it has not been demonstrated for in-situ reactor applications. In this thesis, finite amplitude wave propagation will be used to measure the microstructural evolution in Al-6061. This is the first demonstration of finite amplitude wave propagation at temperatures in excess of 200 °C and during an irradiation test. Second, a method based on contact nonlinear acoustic theory will be developed to identify compressed cracks. Compressed cracks are typically transparent to ultrasonic wave propagation; however, by measuring harmonic content developed during finite amplitude wave propagation, it is shown that even compressed cracks can be characterized. Lastly, piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts

  20. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1970-01-01

    Advances in Nuclear Science and Technology, Volume 5 presents the underlying principles and theory, as well as the practical applications of the advances in the nuclear field. This book reviews the specialized applications to such fields as space propulsion.Organized into six chapters, this volume begins with an overview of the design and objective of the Fast Flux Test Facility to provide fast flux irradiation testing facilities. This text then examines the problem in the design of nuclear reactors, which is the analysis of the spatial and temporal behavior of the neutron and temperature dist

  1. Advanced light water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Giedraityte, Zivile [Helsinki University of Technology, Otaranta 8D-84, 02150 Espoo (Finland)

    2008-07-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  2. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  3. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount

  4. Advanced Measuring (Instrumentation Methods for Nuclear Installations: A Review

    Directory of Open Access Journals (Sweden)

    Wang Qiu-kuan

    2012-01-01

    Full Text Available The nuclear technology has been widely used in the world. The research of measurement in nuclear installations involves many aspects, such as nuclear reactors, nuclear fuel cycle, safety and security, nuclear accident, after action, analysis, and environmental applications. In last decades, many advanced measuring devices and techniques have been widely applied in nuclear installations. This paper mainly introduces the development of the measuring (instrumentation methods for nuclear installations and the applications of these instruments and methods.

  5. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, October 1, 1979-December 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-18

    This report presents the results of work performed from October 1, 1979 through December 31, 1979. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described. This includes: screening creep results, weight gain and post-exposure mechanical properties for materials thermally exposed at 750/sup 0/ and 850/sup 0/C (1382/sup 0/ and 1562/sup 0/F). In addition, the status of the data management system is described.

  6. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1978--March 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-06-26

    The activities associated with the procurement of the materials for the screening test program, information from vendor certification for the materials received, and preliminary information from the materials characterization tests performed by GE are reported. The construction status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment are discussed. The final recommended impurity levels for the screening phase helium are presented and the rational behind this gas chemistry is discussed. The status of the data management system is presented.

  7. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  8. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1972-01-01

    Advances in Nuclear Science and Technology, Volume 6 provides information pertinent to the fundamental aspects of nuclear science and technology. This book covers a variety of topics, including nuclear steam generator, oscillations, fast reactor fuel, gas centrifuge, thermal transport system, and fuel cycle.Organized into six chapters, this volume begins with an overview of the high standards of technical safety for Europe's first nuclear-propelled merchant ship. This text then examines the state of knowledge concerning qualitative results on the behavior of the solutions of the nonlinear poin

  9. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1976-01-01

    Advances in Nuclear Science and Technology, Volume 9 provides information pertinent to the fundamental aspects of nuclear science and technology. This book discusses the safe and beneficial development of land-based nuclear power plants.Organized into five chapters, this volume begins with an overview of the possible consequences of a large-scale release of radioactivity from a nuclear reactor in the event of a serious accident. This text then discusses the extension of conventional perturbation techniques to multidimensional systems and to high-order approximations of the Boltzmann equation.

  10. Development of essential system technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others

    1999-03-01

    Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

  11. An overview of future sustainable nuclear power reactors

    Directory of Open Access Journals (Sweden)

    Andreas Poullikkas

    2013-01-01

    Full Text Available In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA. In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will

  12. Development of inherent core technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  13. Advanced CANDU reactor pre-licensing progress

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2005-07-01

    The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

  14. Metal fire implications for advanced reactors. Part 1, literature review.

    Energy Technology Data Exchange (ETDEWEB)

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.

    2007-10-01

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

  15. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future lunar and Mars robotic and manned missions impose new and innovative technological requirements for their control and protection...

  16. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, April 1, 1982-June 30, 1982

    Energy Technology Data Exchange (ETDEWEB)

    1982-10-01

    Work covered in this report includes the activities associated with the status of the simulated reactor helium supply systems and testing equipment. The progress in the screening test program is described; this includes screening creep results and metallographic analyses for materials thermally exposed or tested at 750/sup 0/, 850/sup 0/, 950/sup 0/ and 1050/sup 0/C (1382/sup 0/, 1562/sup 0/, 1742/sup 0/ and 1922/sup 0/F) in controlled-purity helium. The status of creep-rupture in controlled-purity helium and air and fatigue testing in controlled-purity helium and air in the intensive screening test program is discussed. The results of metallographic studies of corrosion pins exposed at 750/sup 0/C for 6000 hours in controlled-purity helium (solid solution strengthened alloys, centrifugally cast alloys and an iron-base oxide dispersion strengthened alloy) are presented and discussed. The results of metallographic studies on the same materials after 10,000 hour exposure in controlled-purity helium at 850/sup 0/C are also presented.

  17. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  18. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  19. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

    This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

  20. Structural integrity of nuclear reactor pressure vessels

    Science.gov (United States)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  1. Reactivity control assembly for nuclear reactor. [LMFBR

    Science.gov (United States)

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  2. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  3. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  4. Thermionic reactors for space nuclear power

    Science.gov (United States)

    Griaznov, Georgii M.; Zhabotinskii, Evgenii E.; Serbin, Victor I.; Zrodnikov, Anatolii V.; Pupko, Victor Ia.; Ponomarev-Stepnoi, Nikolai N.; Usov, V. A.; Nikolaev, Iu. V.

    Compact thermionic nuclear reactor systems with satisfactory mass performance are competitive with space nuclear power systems based on the organic Rankine and closed Brayton cycles. The mass characteristics of the thermionic space nuclear power system are better than that of the solar power system for power levels beyond about 10 kWe. Longlife thermionic fuel element requirements, including their optimal dimensions, and common requirements for the in-core thermionic reactor design are formulated. Thermal and fast in-core thermionic reactors are considered and the ranges of their sensible use are discussed. Some design features of the fast in-core thermionic reactors cores (power range to 1 MWe) including a choice of coolants are discussed. Mass and dimensional performance for thermionic nuclear power reactor system are assessed. It is concluded that thermionic space nuclear power systems are promising power supplies for spacecrafts and that a single basic type of thermionic fuel element may be used for power requirements ranging to several hundred kWe.

  5. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  6. The DOE advanced gas reactor fuel development and qualification program

    Science.gov (United States)

    Petti, David; Maki, John; Hunn, John; Pappano, Pete; Barnes, Charles; Saurwein, John; Nagley, Scott; Kendall, Jim; Hobbins, Richard

    2010-09-01

    The high outlet temperatures and high thermal-energy conversion efficiency of modular high-temperature gas-cooled reactors (HTGRs) enable an efficient and cost-effective integration of the reactor system with non-electricity-generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The U.S. Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. An overview of the program and recent progress is presented.

  7. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  8. Oklo reactors and implications for nuclear science

    CERN Document Server

    Davis, E D; Sharapov, E I

    2014-01-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross sections are input to all Oklo modeling and we discuss a parameter, the $^{175}$Lu ground state cross section for thermal neutron capture leading to the isomer $^{176\\mathrm{m}}$ Lu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant $\\alpha$ and the ratio $X_q=m_q/\\Lambda$ (where $m_...

  9. Advanced Small Modular Reactor Economics Model Development

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J [ORNL

    2014-10-01

    The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the

  10. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    -plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content.

  11. The role of nuclear reactors in space exploration and development

    Energy Technology Data Exchange (ETDEWEB)

    Lipinski, R.J.

    2000-07-01

    The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. One reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new

  12. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  13. Nuclear Data and the Oklo Natural Nuclear Reactors

    Science.gov (United States)

    Gould, C. R.; Sharapov, E. I.; Sonzogni, A. A.

    2014-04-01

    Data from the Oklo natural nuclear reactors have enabled some of the most sensitive terrestrial tests of time variation of dimensionless fundamental constants. The constraints on variation of αEM, the fine structure constant are particular good, but depend on the reliability of the nuclear data, and on the reliability of the modeling of the reactor environment. We briefly review the history of these tests and discuss our recent work in 1) attempting to better bound the temperatures at which the reactors operated, 2) investigating whether the γ-ray fluxes in the reactors could have contributed to changing lutetium isotopic abundances and 3) determining whether lanthanum isotopic data could provide an alternate estimate of the neutron fluence.

  14. Nuclear reactor alignment plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  15. Actinide transmutation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bultman, J.H.

    1995-01-17

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP).

  16. Advanced Catalytic Hydrogenation Retrofit Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  17. Current Abstracts Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bales, J.D.; Hicks, S.C. [eds.

    1993-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  18. Issues and future direction of thermal-hydraulics research and development in nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P., E-mail: pradip.saha@ge.com [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Aksan, N. [GRNSPG Group, University of Pisa (Italy); Andersen, J. [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Yan, J. [Westinghouse Electric Co., Columbia, SC (United States); Simoneau, J.P. [AREVA, Lyon (France); Leung, L. [Atomic Energy of Canada Ltd., Chalk River, Ontario (Canada); Bertrand, F. [CEA, DEN, DER, F-13108 Saint-Paul-Lez-Durance (France); Aoto, K.; Kamide, H. [Japan Atomic Energy Agency, Chiyoda-ku, Tokyo (Japan)

    2013-11-15

    The paper archives the proceedings of an expert panel discussion on the issues and future direction of thermal-hydraulic research and development in nuclear power reactors held at the NURETH-14 conference in Toronto, Canada, in September 2011. Thermal-hydraulic issues related to both operating and advanced reactors are presented. Advances in thermal-hydraulics have significantly improved the performance of operating reactors. Further thermal-hydraulics research and development is continuing in both experimental and computational areas for operating reactors, reactors under construction or ready for near-term deployment, and advanced Generation-IV reactors. As the computing power increases, the fine-scale multi-physics computational models, coupled with the systems analysis code, are expected to provide answers to many challenging problems in both operating and advanced reactor designs.

  19. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  20. FUEL COMPOSITION FOR NUCLEAR REACTORS

    Science.gov (United States)

    Andersen, J.C.

    1963-08-01

    A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

  1. Cold nuclear fusion reactor and nuclear fusion rocket

    Directory of Open Access Journals (Sweden)

    Huang Zhenqiang

    2013-10-01

    Full Text Available "Nuclear restraint inertial guidance directly hit the cold nuclear fusion reactor and ion speed dc transformer" [1], referred to as "cold fusion reactor" invention patents, Chinese Patent Application No. CN: 200910129632.7 [2]. The invention is characterized in that: at room temperature under vacuum conditions, specific combinations of the installation space of the electromagnetic field, based on light nuclei intrinsic magnetic moment and the electric field, the first two strings of the nuclei to be bound fusion on the same line (track of. Re-use nuclear spin angular momentum vector inherent nearly the speed of light to form a super strong spin rotation gyro inertial guidance features, to overcome the Coulomb repulsion strong bias barrier to achieve fusion directly hit. Similar constraints apply nuclear inertial guidance mode for different speeds and energy ion beam mixing speed, the design of ion speed dc transformer is cold fusion reactors, nuclear fusion engines and such nuclear power plants and power delivery systems start important supporting equipment, so apply for a patent merger

  2. A Nuclear Reactor Transient Methodology Based on Discrete Ordinates Method

    Directory of Open Access Journals (Sweden)

    Shun Zhang

    2014-01-01

    Full Text Available With the rapid development of nuclear power industry, simulating and analyzing the reactor transient are of great significance for the nuclear safety. The traditional diffusion theory is not suitable for small volume or strong absorption problem. In this paper, we have studied the application of discrete ordinates method in the numerical solution of space-time kinetics equation. The fully implicit time integration was applied and the precursor equations were solved by analytical method. In order to improve efficiency of the transport theory, we also adopted some advanced acceleration methods. Numerical results of the TWIGL benchmark problem presented demonstrate the accuracy and efficiency of this methodology.

  3. Development of advanced strain diagnostic techniques for reactor environments.

    Energy Technology Data Exchange (ETDEWEB)

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  4. Medical Radioisotopes Production Without A Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Van der Keur, H.

    2010-05-15

    This report is answering the key question: Is it possible to ban the use of research reactors for the production of medical radioisotopes? Chapter 2 offers a summarized overview on the history of nuclear medicine. Chapter 3 gives an overview of the basic principles and understandings of nuclear medicine. The production of radioisotopes and its use in radiopharmaceuticals as a tracer for imaging particular parts of the inside of the human body (diagnosis) or as an agent in radiotherapy. Chapter 4 lists the use of popular medical radioisotopes used in nuclear imaging techniques and radiotherapy. Chapter 5 analyses reactor-based radioisotopes that can be produced by particle accelerators on commercial scale, other alternatives and the advantages of the cyclotron. Chapter 6 gives an overview of recent developments and prospects in worldwide radioisotopes production. Chapter 7 presents discussion, conclusions and recommendations, and is answering the abovementioned key question of this report: Is it possible to ban the use of a nuclear reactor for the production of radiopharmaceuticals? Is a safe and secure production of radioisotopes possible?.

  5. Testing piezoelectric sensors in a nuclear reactor environment

    Science.gov (United States)

    Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

    2017-02-01

    Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

  6. Advances in nuclear physics

    CERN Document Server

    Vogt, Erich

    1975-01-01

    Review articles on three topics of considerable current interest make up the present volume. The first, on A-hypernuclei, was solicited by the editors in order to provide nuclear physicists with a general description of the most recent developments in a field which this audience has largely neglected or, perhaps, viewed as a novelty in which a bizarre nuclear system gave some information about the lambda-nuclear intersection. That view was never valid. The very recent developments reviewed here-particularly those pertaining to hypernuclear excitations and the strangeness exchange reactions-emphasize that this field provides important information about the models and central ideas of nuclear physics. The off-shell behavior of the nucleon-nucleon interaction is a topic which was at first received with some embarrassment, abuse, and neglect, but it has recently gained proper attention in many nuclear problems. Interest was first focused on it in nuclear many-body theory, but it threatened nuclear physicists'comf...

  7. Wire core reactor for nuclear thermal propulsion

    Science.gov (United States)

    Harty, Richard B.; Brengle, Robert G.

    1993-01-01

    Studies have been performed of a compact high-performance nuclear rocket reactor that incorporates a tungsten alloy wire fuel element. This reactor, termed the wire core reactor, can deliver a specific impulse of 1,000 s using an expander cycle and a nozzle expansion ratio of 500 to 1. The core is constructed of layers of 0.8-mm-dia fueled tungsten wires wound over alternate layers of spacer wires, which forms a rugged annular lattice. Hydrogen flow in the core is annular, flowing from inside to outside. In addition to the concepts compact size and good heat transfer, the core has excellent power-flow matching features and can resist vibration and thermal stresses during star-up and shutdown.

  8. An overview of future sustainable nuclear power reactors

    OpenAIRE

    Andreas Poullikkas

    2013-01-01

    In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are ...

  9. Muon trackers for imaging a nuclear reactor

    Science.gov (United States)

    Kume, N.; Miyadera, H.; Morris, C. L.; Bacon, J.; Borozdin, K. N.; Durham, J. M.; Fuzita, K.; Guardincerri, E.; Izumi, M.; Nakayama, K.; Saltus, M.; Sugita, T.; Takakura, K.; Yoshioka, K.

    2016-09-01

    A detector system for assessing damage to the cores of the Fukushima Daiichi nuclear reactors by using cosmic-ray muon tomography was developed. The system consists of a pair of drift-tube tracking detectors of 7.2× 7.2-m2 area. Each muon tracker consists of 6 x-layer and 6 y-layer drift-tube detectors. Each tracker is capable of measuring muon tracks with 12 mrad angular resolutions, and is capable of operating under 50-μ Sv/h radiation environment by removing gamma induced background with a novel time-coincidence logic. An estimated resolution to observe nuclear fuel debris at Fukushima Daiichi is 0.3 m when the core is imaged from outside the reactor building.

  10. Reference Neutron Radiographs of Nuclear Reactor Fuel

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

  11. Some views on nuclear reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Tanguy, P.Y. [Electricite de France, Paris (France)

    1995-04-01

    This document is the text of a speech given by Pierre Y. Tanguy (Electricite de France) at the 22nd Water Reactor Safety Meeting held in Bethesda, MD in 1994. He describes the EDF nuclear program in broad terms and proceeds to discuss operational safety results with EDF plants. The speaker also outlines actions to enhance safety planned for the future, and he briefly mentions French cooperation with the Chinese on the Daya Bay project.

  12. Multivariable Feedback Control of Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Rune Moen

    1982-07-01

    Full Text Available Multivariable feedback control has been adapted for optimal control of the spatial power distribution in nuclear reactor cores. Two design techniques, based on the theory of automatic control, were developed: the State Variable Feedback (SVF is an application of the linear optimal control theory, and the Multivariable Frequency Response (MFR is based on a generalization of the traditional frequency response approach to control system design.

  13. Nuclear vapor thermal reactor propulsion technology

    Science.gov (United States)

    Maya, Isaac; Diaz, Nils J.; Dugan, Edward T.; Watanabe, Yoichi; McClanahan, James A.; Wen-Hsiung Tu, Carman, Robert L.

    1993-01-01

    The conceptual design of a nuclear rocket based on the vapor core reactor is presented. The Nuclear Vapor Thermal Rocket (NVTR) offers the potential for a specific impulse of 1000 to 1200 s at thrust-to-weight ratios of 1 to 2. The design is based on NERVA geometry and systems with the solid fuel replaced by uranium tetrafluoride (UF4) vapor. The closed-loop core does not rely on hydrodynamic confinement of the fuel. The hydrogen propellant is separated from the UF4 fuel gas by graphite structure. The hydrogen is maintained at high pressure (˜100 atm), and exits the core at 3,100 K to 3,500 K. Zirconium carbide and hafnium carbide coatings are used to protect the hot graphite from the hydrogen. The core is surrounded by beryllium oxide reflector. The nuclear reactor core has been integrated into a 75 klb engine design using an expander cycle and dual turbopumps. The NVTR offers the potential for an incremental technology development pathway to high performance gas core reactors. Since the fuel is readily available, it also offers advantages in the initial cost of development, as it will not require major expenditures for fuel development.

  14. Oklo reactors and implications for nuclear science

    Science.gov (United States)

    Davis, E. D.; Gould, C. R.; Sharapov, E. I.

    2014-04-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross-sections are input to all Oklo modeling and we discuss a parameter, the 175Lu ground state cross-section for thermal neutron capture leading to the isomer 176mLu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant α and the ratio Xq = mq/Λ (where mq is the average of the u and d current quark masses and Λ is the mass scale of quantum chromodynamics (QCD)). We suggest a formula for the combined sensitivity to α and Xq that exhibits the dependence on proton number Z and mass number A, potentially allowing quantum electrodynamic (QED) and QCD effects to be disentangled if a broader range of isotopic abundance data becomes available.

  15. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  16. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    OpenAIRE

    Djurcic, Z.(Argonne National Laboratory, Argonne, Illinois, 60439, U.S.A.); Detwiler, J. A.; Piepke, A.; Foster Jr., V. R.; Miller, L.; Gratta, G.

    2008-01-01

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

  17. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A., E-mail: anlafuente@etsii.upm.e [ETSII-UPM, c/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Piera, M. [ETSII:UNED, c/Juan del Rosal, 12, 28040 Madrid (Spain)

    2010-06-15

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F{sub 2}Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F{sub 2}Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void

  18. Designed porosity materials in nuclear reactor components

    Science.gov (United States)

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  19. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  20. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  1. Johnson Noise Thermometry for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

    2012-09-15

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  2. Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems

    Science.gov (United States)

    2004-12-01

    contrasted with nuclear thermal rockets which use the heat from a nuclear fission reactor to heat propellant to provide rocket thrust and radioisotope...K. Note that the highest temperature (2550 K by the Pewee reactor) was for a nuclear thermal rocket application and has the shortest duration (40 min

  3. Summary of space nuclear reactor power systems, 1983--1992

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  4. Exploring new coolants for nuclear breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lafuente, A. [ETSI Industriales-Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal, 2. 28006 Madrid (Spain)

    2010-07-01

    Breeder reactors are considered the unique tool for fully exploiting the natural nuclear resources. In current LWR, only a 0.5% of the primary energy contained in the nuclei removed from the mine is converted into useful heat, with the rest remaining in the depleted uranium or in the spent fuel. The objective of resource-efficiency stimulated the interest in Fast- Reactor-based fuel cycles which can exploit a much higher fraction of the energy content of the mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles would also offers several potential advantages over a uranium fuel cycle. The coolant initially chosen for most of the FBR programs launched in the 60's was sodium, which still is considered the best candidate for these reactors. However, Na-cooled FBR have a positive void reactivity coefficient, which has been among others, a fundamental drawback that has cancelled the deployment of these reactors. Therefore, it seems reasonable to explore totally new options on coolants for breeders. In this paper, a proposal is presented on a new molten salt (F{sub 2}Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would need an extensive R and D programme, this paper presents the very appealing properties of this salt, in the case of using a specific type of fuel, similar to that of pebble bed reactors. The concept will be studied over a typical MOX composition and extended to a Thorium-based cycle. The general analysis takes into account requirements for criticality (opening the option of hybrid subcritical systems); requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window is found in the definition of a F{sub 2}Be cooled reactor where the safety requirement is met, unlike for molten metal cooled reactors which always have positive void

  5. Recent advances in nuclear cardiology

    DEFF Research Database (Denmark)

    Gutte, H.; Petersen, C. Leth; Kjaer, A.

    2008-01-01

    Nuclear cardiology is an essential part of functional, non-invasive, cardiac imaging. Significant advances have been made in nuclear cardiology since planar (201)thallium ((201)TI) scintigraphy was introduced for the evaluation of left ventricular (LV) perfusion nearly 40 years ago. The use...... of nuclear cardiology has been steadily increasing over the last 20 years with important steps being the introduction of (99m)technetium- ((99m)Tc)-labelled perfusion radiotracers, the change from only planar to now much more single photon emission computed tomography (SPECT) and positron emission tomography...... (PET), electrocardiogram gating of nuclear perfusion imaging, and finally introducing nuclear hybrid imaging using either SPECT or PET together with either computed tomography or magnetic resonance imaging. The indications have extended from nearly only coronary artery diseases to several non...

  6. Dielectric Heaters for Testing Spacecraft Nuclear Reactors

    Science.gov (United States)

    Sims, William Herbert; Bitteker, Leo; Godfroy, Thomas

    2006-01-01

    A document proposes the development of radio-frequency-(RF)-driven dielectric heaters for non-nuclear thermal testing of the cores of nuclear-fission reactors for spacecraft. Like the electrical-resistance heaters used heretofore for such testing, the dielectric heaters would be inserted in the reactors in place of nuclear fuel rods. A typical heater according to the proposal would consist of a rod of lossy dielectric material sized and shaped like a fuel rod and containing an electrically conductive rod along its center line. Exploiting the dielectric loss mechanism that is usually considered a nuisance in other applications, an RF signal, typically at a frequency .50 MHz and an amplitude between 2 and 5 kV, would be applied to the central conductor to heat the dielectric material. The main advantage of the proposal is that the wiring needed for the RF dielectric heating would be simpler and easier to fabricate than is the wiring needed for resistance heating. In some applications, it might be possible to eliminate all heater wiring and, instead, beam the RF heating power into the dielectric rods from external antennas.

  7. Nuclear renaissance in the reactor training of Areva

    Energy Technology Data Exchange (ETDEWEB)

    De Braquilanges, Bertrand [Reactor Training Center/France Manager, La Tour Areva - 1, place Jean Millier - 92084 Paris - La Defense (France); Napior, Amy [Reactor Training Center/USA Manager, 1300 Old Graves Mill Road - Lynchburg VA, 2450 (United States); Schoenfelder, Christian [Reactor Training Center/Germany Manager, Kaiserleistrasse 29 - 63067 Offenbach (Germany)

    2010-07-01

    Because of the perspectives of new builds, a significant increase in the number of design, construction and management personnel working in AREVA, their clients and sub-contractors has been estimated for the next future. In order to cope with the challenge to integrate newly hired people quickly and effectively into the AREVA workforce, a project - 'Training Task Force (TTF)' - was launched in 2008. The objective was to develop introductory and advanced courses and related tools harmonized between AREVA Training Centers in France, Germany and USA. First, a Global Plants Introductory Session (GPIS) was developed for newly hired employees. GPIS is a two weeks training course introducing in a modular way AREVA and specifically the activities and the reactors technical basics. As an example, design and operation of a nuclear power plant is illustrated on EPRTM. Since January 2009, these GPIS are held regularly in France, Germany and the US with a mixing of employees from these 3 regions. Next, advanced courses for more experienced employees were developed: - Advanced EPR{sup TM}, giving a detailed presentation of the EPR{sup TM} reactor design; - Codes and Standards; - Technical Nuclear Safety. Finally, feasibility studies on a Training Material Management (TMM) system, able to manage the training documentation, and on a worldwide training administration tool, were performed. The TTF project was completed mid of 2009; it transferred their recurrent activities to a new AREVA training department. This unit now consists of the French, German and US Reactors Training Centers. In particular, all courses developed by the TTF are now implemented worldwide with an opening to external trainees. The current worldwide course catalogue includes training courses for operation and maintenance personnel as well as for managers, engineers and non technical personnel of nuclear operators, suppliers, safety authorities and expert organizations. Training delivery is supported

  8. Fuel, Structural Material and Coolant for an Advanced Fast Micro-Reactor

    Science.gov (United States)

    Do Nascimento, J. A.; Duimarães, L. N. F.; Ono, S.

    The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials.

  9. Solution of heat removal from nuclear reactors by natural convection

    Directory of Open Access Journals (Sweden)

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  10. Parameter Identification with the Random Perturbation Particle Swarm Optimization Method and Sensitivity Analysis of an Advanced Pressurized Water Reactor Nuclear Power Plant Model for Power Systems

    Directory of Open Access Journals (Sweden)

    Li Wang

    2017-02-01

    Full Text Available The ability to obtain appropriate parameters for an advanced pressurized water reactor (PWR unit model is of great significance for power system analysis. The attributes of that ability include the following: nonlinear relationships, long transition time, intercoupled parameters and difficult obtainment from practical test, posed complexity and difficult parameter identification. In this paper, a model and a parameter identification method for the PWR primary loop system were investigated. A parameter identification process was proposed, using a particle swarm optimization (PSO algorithm that is based on random perturbation (RP-PSO. The identification process included model variable initialization based on the differential equations of each sub-module and program setting method, parameter obtainment through sub-module identification in the Matlab/Simulink Software (Math Works Inc., Natick, MA, USA as well as adaptation analysis for an integrated model. A lot of parameter identification work was carried out, the results of which verified the effectiveness of the method. It was found that the change of some parameters, like the fuel temperature and coolant temperature feedback coefficients, changed the model gain, of which the trajectory sensitivities were not zero. Thus, obtaining their appropriate values had significant effects on the simulation results. The trajectory sensitivities of some parameters in the core neutron dynamic module were interrelated, causing the parameters to be difficult to identify. The model parameter sensitivity could be different, which would be influenced by the model input conditions, reflecting the parameter identifiability difficulty degree for various input conditions.

  11. Collective control of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rognin, L.

    1995-06-01

    Nowadays, mainly related to the increasing complexity of working environments, working activities become more and collective. The present research on the paradoxical nature of working teams, considered from a reliability point of view. This document is composed of four Sections. The first Section introduces the context of the research, its objectives and the underlying assumptions. In the second Section, we describe a working situation, which is the control of a nuclear reactor. Relations between cooperative work and reliability are discussed in the third Section. Finally, in the fourth Section, a synthesis of the research and some perspectives are proposed. (authors). 7 refs.

  12. Mirror Advanced Reactor Study interim design report

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  13. Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

  14. The MAUS nuclear space reactor with ion propulsion system

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, Enrico [DINCE - Dipartimento di Ingegneria Nucleare e Conversioni Energetiche, University of Rome ' La Sapienza' , C.so V. Emanuele II, 244, 00186 Rome (Italy)]. E-mail: mainardi@frascati.enea.it

    2006-06-01

    MAUS (Moltiplicatore Avanzato Ultracompatto Spaziale) is a nuclear reactor concept design capable to ensure a reliable, long-lasting, low-mass, compact energy supply needed for advanced, future space missions. The exploration of the solar system and the space beyond requires the development of nuclear energy generators for supplying electricity to space-bases, spacecrafts, probes or satellites, as well as for propelling ships in long space missions. For propulsion, the MAUS nuclear reactor could be used to power electric ion drive engines. An ion engine is able to build up to very high velocities, far greater than chemical propulsion systems, but has high power and long service requirements. The MAUS concept is described, together with the ion propulsion engine and together with the reference thermoionic process used to convert the thermal power into electricity. The design work has been performed at the Nuclear Engineering and Energy Conversion Department of the University of Rome 'La Sapienza' starting from 1992 on an issue submitted by the Italian Space Agency (ASI), in cooperation with the research laboratories of ENEA.

  15. The Maus nuclear space reactor with ion propulsion system

    Energy Technology Data Exchange (ETDEWEB)

    Enrico Mainardi [DINCE - Dipartimento di Ingegneria Nucleare e Conversioni Energetiche, University of Rome ' La Sapienza' , C.so V. EmanueleII, 244, 00186 Roma (Italy)

    2006-07-01

    MAUS (Moltiplicatore Avanzato Ultracompatto Spaziale) is a nuclear reactor concept design capable to ensure a reliable, long lasting, low mass, compact energy supply needed for advanced, future space missions. The exploration of the solar system and the space beyond requires the development of nuclear energy generators for supplying electricity to space-bases, spacecrafts, probes or satellites, as well as for propelling ships in long space missions. For propulsion, the MAUS nuclear reactor could be used to power electric ion drive engines. An ion engine is able to build up to very high velocities, far greater than chemical propulsion systems, but has high power and long service requirements. The MAUS concept is described, together with the ion propulsion engine and together with the reference thermionic process used to convert the thermal power into electricity. The design work has been performed at the Nuclear Engineering and Energy Conversion Department of the University of Rome 'La Sapienza' starting from 1992 on an issue submitted by the Italian Space Agency (ASI), in cooperation with the research laboratories of ENEA. (author)

  16. The use of PRA in the development of ALWR (advanced light water reactor) design requirements. [Advanced Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Summitt, R.L. (Safety and Reliability Optimization Services, Inc., Knoxville, TN (USA)); Additon, S.L. (TENERA, L.P., Bethesda, MD (USA)); Pasedag, W.F. (USDOE, Washington, DC (USA))

    1989-01-01

    The current hiatus in nuclear power plant orders provides an opportunity for the development of advanced light water reactor (ALWR) design concepts and regulatory requirements which incorporate the insights gained from the application of the probabilistic risk assessment. The US Department of Energy is assisting the Electric Power Research Institute (EPRI) in the incorporation of PRA insights into the specification of the utility requirements, and reactor vendors in support of the conceptual design of safety systems, for such advanced plants. This paper reviews the applications of PRA methods in this development of specifications for, and the design of simplified, rugged ALWRs with a significantly improved risk profile. Specific examples of the impact of utilizing published PRA insights, construction and use of functional PRA models, and feedback of PRA experience into the specification of the key assumptions and groundrules for ALWR PRAs are presented. 13 refs., 3 tabs.

  17. Advanced nuclear precleaner

    Energy Technology Data Exchange (ETDEWEB)

    Wright, S.R. [InnovaTech, Inc., Durham, NC (United States)

    1997-10-01

    This Phase II Small Business Innovation Research (SBIR) program`s goal is to develop a dynamic, self-cleaning air precleaner for high-efficiency particulate air (HEPA) filtration systems that would extend significantly the life of HEPA filter banks by reducing the particulate matter that causes filter fouling and increased pack pressure. HEPA filters are widely used in DOE, Department of Defense, and a variety of commercial facilities. InnovaTech, Inc. (Formerly Micro Composite materials Corporation) has developed a proprietary dynamic separation device using a concept called Boundary Layer Momentum Transfer (BLMT) to extract particulate matter from fluid process streams. When used as a prefilter in the HVAC systems or downstream of waste vitrifiers in nuclear power plants, fuel processing facilities, and weapons decommissioning factories, the BLMT filter will dramatically extend the service life and increase the operation efficiency of existing HEPA filtration systems. The BLMT filter is self cleaning, so there will be no degraded flow or increased pressure drop. Because the BLMT filtration process is independent of temperature, it can be designed to work in ambient, medium, or high-temperature applications. During Phase II, the authors are continuing development of the computerized flow simulation model to include turbulence and incorporate expansion into a three-dimensional model that includes airflow behavior inside the filter housing before entering the active BLMT device. A full-scale (1000 ACFM) prototype filter is being designed to meet existing HEPA filter standards and will be fabricated for subsequent testing. Extensive in-house testing will be performed to determine a full range of performance characteristics. Final testing and evaluation of the prototype filter will be conducted at a DOE Quality Assurance Filter Test Station.

  18. Assessment of Sensor Technologies for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, Kofi [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vlim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Kisner, Roger A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Britton, Jr, Charles L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wootan, D. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anheier, Jr, N. C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Diaz, A. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, E. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chien, H. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Sheen, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States); Gopalsami, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Heifetz, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Tam, S. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Upadhyaya, B. R. [Univ. of Tennessee, Knoxville, TN (United States); Stanford, A. [Univ. of Tennessee, Knoxville, TN (United States)

    2016-10-01

    Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributes to the design and implementation of AdvRx concepts.

  19. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    Energy Technology Data Exchange (ETDEWEB)

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  1. Advanced fuels for thermal spectrum reactors

    OpenAIRE

    Zakova, Jitka

    2012-01-01

    The advanced fuels investigated in this thesis comprise fuels non− conventional in their design/form (TRISO), their composition (high content of plutonium and minor actinides) or their use in a reactor type, in which they have not been used before (e.g. nitride fuel in BWR). These fuels come with a promise of improved characteristics such as safe, high temperature operation, spent fuel transmutation or fuel cycle extension, for which reasons their potentialis worth assessment and investigatio...

  2. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  3. Johnson Noise Thermometry for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Britton Jr, Charles L [ORNL; Roberts, Michael [ORNL; Bull, Nora D [ORNL; Holcomb, David Eugene [ORNL; Wood, Richard Thomas [ORNL

    2012-10-01

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  4. Recent advances in nuclear cardiology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Woo [Dept. Nuclear Medicine, Seoul National University Bundang Hospital, Seoul National University College of Medicine, Seongnam (Korea, Republic of)

    2016-09-15

    Nuclear cardiology is one of the major fields of nuclear medicine practice. Myocardial perfusion studies using single-photon emission computed tomography (SPECT) have played a crucial role in the management of coronary artery diseases. Positron emission tomography (PET) has also been considered an important tool for the assessment of myocardial viability and perfusion. However, the recent development of computed tomography (CT)/magnetic resonance imaging (MRI) technologies and growing concerns about the radiation exposure of patients remain serious challenges for nuclear cardiology. In response to these challenges, remarkable achievements and improvements are currently in progress in the field of myocardial perfusion imaging regarding the applicable software and hardware. Additionally, myocardial perfusion positron emission tomography (PET) is receiving increasing attention owing to its unique capability of absolute myocardial blood flow estimation. An F-18-labeled perfusion agent for PET is under clinical trial with promising interim results. The applications of F-18 fluorodeoxyglucose (FDG) and F-18 sodium fluoride (NaF) to cardiovascular diseases have revealed details on the basic pathophysiology of ischemic heart diseases. PET/MRI seems to be particularly promising for nuclear cardiology in the future. Restrictive diseases, such as cardiac sarcoidosis and amyloidosis, are effectively evaluated using a variety of nuclear imaging tools. Considering these advances, the current challenges of nuclear cardiology will become opportunities if more collaborative efforts are devoted to this exciting field of nuclear medicine.

  5. Development of an automated core model for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R.D.

    1998-12-31

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    Energy Technology Data Exchange (ETDEWEB)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-03-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized.

  8. ANALISIS TRANSIEN PADA FIXED BED NUCLEAR REACTOR

    Directory of Open Access Journals (Sweden)

    M. Rizaal

    2015-03-01

    Full Text Available Desain teras Fixed Bed Nuclear Reactor (FBNR yang modular memungkinkan pengendalian daya dapat dilakukan dengan mengatur ketinggian suspended core dan laju aliran massa pendingin. Tujuan penelitian ini adalah mempelajari perubahan daya termal teras sebagai akibat perubahan laju aliran massa pendingin yang masuk ke teras reaktor dan perubahan ketinggian suspended core serta mempelajari karakteristik keselamatan melekat yang dimiliki FBNR saat terjadi kegagalan pelepasan kalor (loss of heat sink. Keadaan neutronik teras dimodelkan pada kondisi tunak dengan menggunakan paket program Standard Reactor Analysis Code (SRAC untuk memperoleh data fluks neutron, konstanta grup, fraksi neutron kasip, konstanta peluruhan prekursor neutron kasip, dan beberapa parameter teras penting lainnya. Selanjutnya data tersebut digunakan pada perhitungan transien sebagai syarat awal. Analisis transien dilakukan pada tiga kondisi, yaitu saat terjadi penurunan laju aliran massa pendingin, saat terjadi penurunan ketinggian suspended core, dan saat terjadi kegagalan sistem pelepasan kalor. Hasil yang diperoleh dari penelitian ini menunjukkan bahwa penurunan laju aliran massa pendingin sebesar 50%, dari kondisi normal, menyebabkan daya termal teras turun 28% dibanding daya sebelumnya. Penurunan ketinggian suspended core sebesar 30% dari ketinggian normal menyebabkan daya termal teras turun 17% dibanding daya sebelumnya. Sementara untuk kondisi kegagalan sistem pelepasan kalor, daya termal teras mengalami penurunan sebesar 76%. Dengan demikian, pengendalian daya pada FBNR dapat dilakukan dengan mengatur laju aliran massa pendingin dan ketinggian suspended core, serta keselamatan melekat yang handal pada kondisi kegagalan sistem pelepasan kalor. Kata kunci: FBNR, transien, daya, laju aliran massa, suspended core Modular in design enables Fixed Bed Nuclear Reactor (FBNR power controlled by the adjustment of suspended core and coolant flow rate. The main purposes of this paper

  9. Hybrid reactors: Nuclear breeding or energy production?

    Energy Technology Data Exchange (ETDEWEB)

    Piera, Mireia [UNED, ETSII-Dp Ingenieria Energetica, c/Juan del Rosal 12, 28040 Madrid (Spain); Lafuente, Antonio; Abanades, Alberto; Martinez-Val, J.M. [ETSII-UPM, c/Jose Gutierrez Abascal 2, 28006 Madrid (Spain)

    2010-09-15

    After reviewing the long-standing tradition on hybrid research, an assessment model is presented in order to characterize the hybrid performance under different objectives. In hybrids, neutron multiplication in the subcritical blanket plays a major role, not only for energy production and nuclear breeding, but also for tritium breeding, which is fundamental requirement in fusion-fission hybrids. All three objectives are better achieved with high values of the neutron multiplication factor (k-eff) with the obvious and fundamental limitation that it cannot reach criticality under any event, particularly, in the case of a loss of coolant accident. This limitation will be very important in the selection of the coolant. Some general considerations will be proposed, as guidelines for assessing the hybrid potential in a given scenario. Those guidelines point out that hybrids can be of great interest for the future of nuclear energy in a framework of Sustainable Development, because they can contribute to the efficient exploitation of nuclear fuels, with very high safety features. Additionally, a proposal is presented on a blanket specially suited for fusion-fission hybrids, although this reactor concept is still under review, and new work is needed for identifying the most suitable blanket composition, which can vary depending on the main objective of the hybrid. (author)

  10. Nuclear reactors built, being built, or planned 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.

  11. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  12. Selection of nuclear reactors through the hierarchic analysis process: the Mexican case; Seleccion de reactores nucleares mediante el proceso de analisis jerarquico: el caso Mexicano

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo, C.; Nelson, P.F.; Francois, J.L. [Departamento de Sistemas Energeticos, Facultad de Ingenieria, UNAM, Paseo Cuauhnahuac 8532, Jiutepec, 62550 Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2008-07-01

    In this work the decision making method known as hierarchical analysis process for the selection of a new reactor in Mexico was applied. The main objective of the process it is to select the nuclear reactor technology more appropriate for Mexico, to begin the bid process inside one or two years to begin their operation in 2016. The options were restricted to four reactors that fulfill the following ones approaches: 1) its are advanced reactors, from the technological point of view, with regard to the reactors that at the moment operate in the Laguna Verde Power Station, 2) its are reactors that have the totally finished design, 3) its are reactors that already have the certification on the part of the regulator organism of the origin country or that they are in an advanced state of the certification process and 4) its are reactors offered by the companies that they have designed and built the greater number of reactors that are at the moment in operation at world level. Taking into account these restrictions it was decided to consider as alternative at the reactors: Advanced Boiling Water Reactor (A BWR), European Reactor of Pressurized Water (EPR), Water at Pressure reactor (AP1000) and Simplified Economic Reactor of Boiling Water (ESBWR). The evaluation approaches include economic and of safety indicators, qualitative some of them and other quantitative ones. Another grade of complexity in the solution of the problem is that there are actors that can be involved in the definition of the evaluation approaches and in the definition of the relative importance among them, according to each actor's interests. To simplify the problem its were only considered two actors or groups of interest that can influence in more significant way and that are the Federal Commission of Electricity and the National Commission of Nuclear Safety and Safeguards. The qualifications for each reactor in function of the evaluation approaches were obtained, being the A BWR the best

  13. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    OpenAIRE

    Vladimir Petrochenko; Georgy Toshinsky

    2012-01-01

    On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral) design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing...

  14. Spent nuclear fuel discharges from U.S. reactors 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  15. Inherently safe reactors and a second nuclear era.

    Science.gov (United States)

    Weinberg, A M; Spiewak, I

    1984-06-29

    The Swedish PIUS reactor and the German-American small modular high-temperature gas-cooled reactor are inherently safe-that is, their safety relies not upon intervention of humans or of electromechanical devices but on immutable principles of physics and chemistry. A second nuclear era may require commercialization and deployment of such inherently safe reactors, even though existing light-water reactors appear to be as safe as other well-accepted sources of central electricity, particularly hydroelectric dams.

  16. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Djurcic, Zelimir; Detwiler, Jason A.; Piepke, Andreas; Foster Jr., Vince R.; Miller, Lester; Gratta, Giorgio

    2008-08-06

    Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in {bar {nu}}{sub e} detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties, and their relevance to reactor {bar {nu}}{sub e} experiments.

  17. Nuclear reactor materials at the atomic scale

    Directory of Open Access Journals (Sweden)

    Emmanuelle A. Marquis

    2009-11-01

    Full Text Available With the renewed interest in nuclear energy, developing new materials able to respond to the stringent requirements of the next-generation fission and future fusion reactors has become a priority. An efficient search for such materials requires detailed knowledge of material behaviour under irradiation, high temperatures and corrosive environments. Minimizing the rates of materials degradation will be possible only if the mechanisms by which it occurs are understood. Atomic-scale experimental probing as well as modelling can provide some answers and help predict in-service behaviour. This article illustrates how this approach has already improved our understanding of precipitation under irradiation, corrosion behaviour, and stress corrosion cracking. It is also now beginning to provide guidance for the development of new alloys.

  18. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  19. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.K.; Snell, V. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca; West, J. [Candesco, Toronto, Ontario (Canada); Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada)

    2006-07-01

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. AECL initiated pre-licensing reviews of the ACR reactor design in Canada, US and China, with an objective to take into account regulatory feedback early in the design process. The Canadian Nuclear Safety Commission (CNSC) is performing a pre-project pre-licensing assessment of the ACR design. The objective of the assessment is to issue a formal statement as to whether there are any fundamental barriers that would prevent the licensing of the new CANDU reactor design in Canada under the Nuclear Safety and Control Act. The CNSC review is being conducted in four phases. In Phase 1 (September 2003 to September 2004) CNSC performed a pre-licensing review of the ACR-700, and focused on the design process, methodology, design concepts and R and D. CNSC staff reviewed about 100 reports, and submitted to AECL questions and comments. In Phase 2 (September 2004 to August 2005) AECL provided responses and additional information to CNSC on their comments and questions in Phase 1. Phase 3 is the Transition Phase (September 2005 to May 2006), bridging the transition from the ACR-700 to the ACR-1000 design. Phase 3 focused on review of generic aspects of the ACR design, on the Safety

  20. Meteodiffusive Characterization of Algiers' Nuclear Research Reactor

    Directory of Open Access Journals (Sweden)

    Mourad Messaci

    2007-01-01

    Full Text Available In the framework of the environmental impact studies of the nuclear research reactor of Algiers, we will present the work related to the atmospheric dispersion of releases due to the installation in normal operation, which dealt with the assessment of spatial distribution of yearly average values of atmospheric dilution factor. The aim of this work is a characterization of the site in terms of diffusivity, which is basic for the radiological impact evaluation of the reactor. The meteorological statistics result from the National Office of Meteorology and concern 15 years of hourly records. According to the nature and features of these data, a Gaussian-type model with wind direction sectors was used. Values of wind speed at release height were estimated from measurement values at 10 m from ground. For the assessment of vertical dispersion coefficient, we used Briggs' formulas related to a sampling time of one hour. Areas of maximum impact were delimited and points of highest concentration within these zones were identified.

  1. Role of research reactors for nuclear power program in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Soentono, S.; Arbie, B. [National Atomic Energy Agency, Batan (Indonesia)

    1994-12-31

    The main objectives of nuclear development program in Indonesia are to master nuclear science and technology, as well as to utilise peaceful uses of nuclear know-how, aiming at stepwisely socioeconomic development. A Triga Mark II, previously of 250 kW, reactor in Bandung has been in operation since 1965 and its design power has been increased to 1000 kW in 1972. Using core grid of the Triga 250 kW, BATAN designed and constructed the Kartini Reactor in Yogyakarta which started its operation in 1979. Both of these Triga reactors have served a wide spectrum of utilisation, such as training of manpower in nuclear engineering as well as radiochemistry, isotope production and beam research activities in solid state physics. In order to support the nuclear power development program in general and to suffice the reactor experiments further, simultaneously meeting the ever increasing demand for radioisotope, the third reactor, a multipurpose reactor of 30 MW called GA. Siwabessy (RSG-GAS) has been in operation since 1987 at Serpong near Jakarta. Each of these reactors has strong cooperation with Universities, namely the Bandung Institute of Technology at Bandung, the Gadjah Mada University at Yogyakarta, and the Indonesia University at Jakarta and has facilitated the man power development required. The role of these reactors, especially the multipurpose GA. Siwabessy reactor, as essential tools in nuclear power program are described including the experience gained during preproject, construction and commissioning, as well as through their operation, maintenance and utilisation.

  2. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  3. Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

    CERN Document Server

    Kessler, Günter

    2012-01-01

    Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

  4. Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

    Energy Technology Data Exchange (ETDEWEB)

    Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Materials Science in Radiation and Dynamics Extremes

    2016-09-26

    In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

  5. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  6. Prognostics Health Management for Advanced Small Modular Reactor Passive Components

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-10-18

    In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

  7. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    King, Jeffrey C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines (CSM), Golden, CO (United States); Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F., E-mail: guimaraes@ieav.cta.br, E-mail: mencarini@ieav.cta.br [Instituto de Estudos Avancados (IEAV), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear

    2015-07-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW{sub e} and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k{sub eff} = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  8. A brief history of design studies on innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  9. A brief history of design studies on innovative nuclear reactors

    Science.gov (United States)

    Sekimoto, Hiroshi

    2014-09-01

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  10. Nuclear reactors built, being built, or planned, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  11. A combined gas cooled nuclear reactor and fuel cell cycle

    Science.gov (United States)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  12. Theory of neutron slowing down in nuclear reactors

    CERN Document Server

    Ferziger, Joel H; Dunworth, J V

    2013-01-01

    The Theory of Neutron Slowing Down in Nuclear Reactors focuses on one facet of nuclear reactor design: the slowing down (or moderation) of neutrons from the high energies with which they are born in fission to the energies at which they are ultimately absorbed. In conjunction with the study of neutron moderation, calculations of reactor criticality are presented. A mathematical description of the slowing-down process is given, with particular emphasis on the problems encountered in the design of thermal reactors. This volume is comprised of four chapters and begins by considering the problems

  13. The current status of nuclear research reactor in Thailand

    Energy Technology Data Exchange (ETDEWEB)

    Sittichai, C.; Kanyukt, R.; Pongpat, P. [Office of Atomic Energy for Peace, Bangkok (Thailand)

    1998-10-01

    Since 1962, the Thai Research Reactor has been serving for various kinds of activities i.e. the production of radioisotopes for medical uses and research and development on nuclear science and technology, for more than three decades. The existing reactor site should be abandoned and relocated to the new suitable site, according to Thai cabinet`s resolution on the 27 December 1989. The decommissioning project for the present reactor as well as the establishment of new nuclear research center were planned. This paper discussed the OAEP concept for the decommissioning programme and the general description of the new research reactor and some related information were also reported. (author)

  14. A study on future nuclear reactor technology and development strategy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. Y.; Kim, S. H.; Sohn, D. S.; Suk, S. D.; Zee, S. K.; Yang, M. H.; Kim, H. J.; Park, W. S

    2000-12-01

    Development of nuclear reactor and fuel cycle technology for future is essential to meet the current issues such as enhancement of nuclear power reactor safety, economically competitive with gas turbine power generation, less production of radioactive waste, proliferation resistant fuel cycle, and public acceptance in consideration of lack of energy resources in the nuclear countries worldwide as well as in Korea. This report deals with as follows, 1) Review the world energy demand and supply perspective and analyse nature of energy and sustainable development to set-up nuclear policy in Korea 2) Recaptitulate the current long term nuclear R and D activities 3) Review nuclear R and D activities and programs of USA, Japan, France, Russia, international organizations such as IAEA, OECD/NEA 4) Recommend development directions of nuclear reactors and fuels.

  15. Pre-licensing of the Advanced CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ion, R.; Popov, N.; Snell, V.; Xu, C. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); West, J. [Candesco Co., Toronto, Ontario (Canada)

    2006-09-15

    Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life, and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross electrical output of 1165 MWe. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU systems, components, and materials, as well as the experience and feedback received from owners and operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The CANDU system is ideally suited to this evolutionary approach since the modular fuel channel reactor design can be modified, through a series of incremental changes in the reactor core design, to increase the power output and improve the overall safety, economics, and performance. The safety enhancements made in ACR-1000 encompass improved safety margins, performance and reliability of safety related systems. In particular, the use of the CANFLEX-ACR fuel bundle, with lower linear rating and higher critical heat flux, provides increased operating and safety margins. Safety features draw from those of the existing CANDU plants (e.g., the two

  16. The necessity of nuclear reactors for targeted radionuclide therapies.

    Science.gov (United States)

    Krijger, Gerard C; Ponsard, Bernard; Harfensteller, Mark; Wolterbeek, Hubert T; Nijsen, Johannes W F

    2013-07-01

    Nuclear medicine has been contributing towards personalized therapies. Nuclear reactors are required for the working horses of both diagnosis and treatment, i.e., Tc-99m and I-131. In fact, reactors will remain necessary to fulfill the demand for a variety of radionuclides and are essential in the expanding field of targeted radionuclide therapies for cancer. However, the main reactors involved in the global supply are ageing and expected to shut down before 2025. Therefore, the fields of (nuclear) medicine, nuclear industry and politics share a global responsibility, faced with the task to secure future access to suitable nuclear reactors. At the same time, alternative production routes should be industrialized. For this, a coordinating entity should be put into place.

  17. Compiled reports on the applicability of selected codes and standards to advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

    1994-08-01

    The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC.

  18. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  19. Advanced Burner Reactor 1000MWth Reference Concept

    Energy Technology Data Exchange (ETDEWEB)

    Cahalan, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fanning, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kellogg, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, L. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Momozaki, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Reed, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Salev, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Seidensticker, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Tang, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Tzanos, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Wei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Chikazawa, Y. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2007-09-30

    The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence, to validate the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat.

  20. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  1. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Science.gov (United States)

    2012-10-12

    ... Section 19.3, ``Regulatory Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water...] [FR Doc No: 2012-25110] NUCLEAR REGULATORY COMMISSION [NRC-2012-0237] Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory...

  2. Advanced Fast Reactor - 100 (AFR-100) Report for the Technical Review Panel

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, Anton [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, Taek K. [Argonne National Lab. (ANL), Argonne, IL (United States); Middleton, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-06-04

    This report is written to provide an overview of the Advanced Fast Reactor-100 in the requested format for a DOE technical review panel. This report was prepared with information that is responsive to the DOE Request for Information, DE-SOL-0003674 Advanced Reactor Concepts, dated February 27, 2012 from DOE’s Office of Nuclear Energy, Office of Nuclear Reactor Technologies. The document consists of two main sections. The first section is a summary of the AFR-100 design including the innovations that are incorporated into the design. The second section contains a series of tables that respond to the various questions requested of the reactor design team from the subject DOE RFI.

  3. Nuclear reactors built, being built, or planned 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.

  4. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  5. Neutron spectrometer for fast nuclear reactors

    CERN Document Server

    Osipenko, M; Ricco, G; Caiffi, B; Pompili, F; Pillon, M; Angelone, M; Verona-Rinati, G; Cardarelli, R; Mila, G; Argiro, S

    2015-01-01

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  6. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  7. Experimental determination of nuclear parameters for RP-0 reactor core; Determinacion experimental de los parametros nucleares para el nucleo tipo MTR del reactor nuclear RP-0

    Energy Technology Data Exchange (ETDEWEB)

    Cajacuri, Rafael A. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    2000-07-01

    In the nuclear reactor for investigations RP-0 which is in Lima, Peru, that is a open pool class reactor with 1 to 10 watts of power and as a nuclear fuel uranium 238 enriched to 20% constituted by elements of Material Testing Reactor fuel class. This has reflectors of graphite and moderator of water demineralized. In 1996/1997 was measured in this reactor the following parameters: position of the control bar that make critic the reactor, critic height of moderator, excess of reactivity of the nucleus, parameter of reactivity for vacuum, parameter of reactivity for temperature, reactivity of its control bar, levels of doses in the reactor. (author)

  8. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  9. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  10. Nuclear reactors built, being built, or planned, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  11. Nuclear reactors built, being built, or planned: 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  12. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    Science.gov (United States)

    Carmack, W. J.; Husser, D. L.; Mohr, T. C.; Richardson, W. C.

    2004-02-01

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  13. Nuclear integrated database and design advancement system

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Jae Joo; Jeong, Kwang Sub; Kim, Seung Hwan; Choi, Sun Young

    1997-01-01

    The objective of NuIDEAS is to computerize design processes through an integrated database by eliminating the current work style of delivering hardcopy documents and drawings. The major research contents of NuIDEAS are the advancement of design processes by computerization, the establishment of design database and 3 dimensional visualization of design data. KSNP (Korea Standard Nuclear Power Plant) is the target of legacy database and 3 dimensional model, so that can be utilized in the next plant design. In the first year, the blueprint of NuIDEAS is proposed, and its prototype is developed by applying the rapidly revolutionizing computer technology. The major results of the first year research were to establish the architecture of the integrated database ensuring data consistency, and to build design database of reactor coolant system and heavy components. Also various softwares were developed to search, share and utilize the data through networks, and the detailed 3 dimensional CAD models of nuclear fuel and heavy components were constructed, and walk-through simulation using the models are developed. This report contains the major additions and modifications to the object oriented database and associated program, using methods and Javascript.. (author). 36 refs., 1 tab., 32 figs.

  14. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  15. Advanced nuclear thermal propulsion concepts

    Science.gov (United States)

    Howe, Steven D.

    1993-11-01

    In 1989, a Presidential directive created the Space Exploration Initiative (SEI) which had a goal of placing mankind on Mars in the early 21st century. The SEI was effectively terminated in 1992 with the election of a new administration. Although the initiative did not exist long enough to allow substantial technology development, it did provide a venue, for the first time in 20 years, to comprehensively evaluate advanced propulsion concepts which could enable fast, manned transits to Mars. As part of the SEI based investigations, scientists from NASA, DoE National Laboratories, universities, and industry met regularly and proceeded to examine a variety of innovative ideas. Most of the effort was directed toward developing a solid-core, nuclear thermal rocket and examining a high-power nuclear electric propulsion system. In addition, however, an Innovative Concepts committee was formed and charged with evaluating concepts that offered a much higher performance but were less technologically mature. The committee considered several concepts and eventually recommended that further work be performed in the areas of gas core fission rockets, inertial confinement fusion systems, antimatter based rockets, and gas core fission electric systems. Following the committee's recommendations, some computational modeling work has been performed at Los Alamos in certain of these areas and critical issues have been identified.

  16. Advanced nuclear thermal propulsion concepts

    Science.gov (United States)

    Howe, Steven D.

    1993-01-01

    In 1989, a Presidential directive created the Space Exploration Initiative (SEI) which had a goal of placing mankind on Mars in the early 21st century. The SEI was effectively terminated in 1992 with the election of a new administration. Although the initiative did not exist long enough to allow substantial technology development, it did provide a venue, for the first time in 20 years, to comprehensively evaluate advanced propulsion concepts which could enable fast, manned transits to Mars. As part of the SEI based investigations, scientists from NASA, DoE National Laboratories, universities, and industry met regularly and proceeded to examine a variety of innovative ideas. Most of the effort was directed toward developing a solid-core, nuclear thermal rocket and examining a high-power nuclear electric propulsion system. In addition, however, an Innovative Concepts committee was formed and charged with evaluating concepts that offered a much higher performance but were less technologically mature. The committee considered several concepts and eventually recommended that further work be performed in the areas of gas core fission rockets, inertial confinement fusion systems, antimatter based rockets, and gas core fission electric systems. Following the committee's recommendations, some computational modeling work has been performed at Los Alamos in certain of these areas and critical issues have been identified.

  17. Construction of the advanced boiling water reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Natsume, Nobuo; Noda, Hiroshi [Tokyo Electric Power Co. (Japan). Nuclear Power Plant Construction Dept.

    1996-07-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7.

  18. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  19. Fundamental Thermal Fluid Physics of High Temperature Flows in Advanced Reactor Systems - Nuclear Energy Research Initiative Program Interoffice Work Order (IWO) MSF99-0254 Final Report for Period 1 August 1999 to 31 December 2002

    Energy Technology Data Exchange (ETDEWEB)

    McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E. (INEEL); Shenoy, A.; Baccaglini, G. (General Atomics); Pletcher, R.H. (Iowa State U.); Wallace, J.M.; Vukoslavcevic, P. (U. Maryland); Jackson, J.D. (U. Manchester, UK); Kunugi, T. (Kyoto U., Japan); Satake, S.-i. (Tokyo U. Science, Japan)

    2002-12-31

    The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of advanced reactors for higher efficiency and enhanced safety and for deployable reactors for electrical power generation, process heat utilization and hydrogen generation. While key applications would be advanced gas-cooled reactors (AGCRs) using the closed Brayton cycle (CBC) for higher efficiency (such as the proposed Gas Turbine - Modular Helium Reactor (GT-MHR) of General Atomics [Neylan and Simon, 1996]), results of the proposed research should also be valuable in reactor systems with supercritical flow or superheated vapors, e.g., steam. Higher efficiency leads to lower cost/kwh and reduces life-cycle impacts of radioactive waste (by reducing waters/kwh). The outcome will also be useful for some space power and propulsion concepts and for some fusion reactor concepts as side benefits, but they are not the thrusts of the investigation. The objective of the project is to provide fundamental thermal fluid physics knowledge and measurements necessary for the development of the improved methods for the applications.

  20. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development program. Progress report, October 1, 1981-December 31, 1981. [Alloy-MA-956; alloy-MA-754

    Energy Technology Data Exchange (ETDEWEB)

    Kimball, O.F.

    1982-06-15

    Work covered in this report includes the activities associated with the status of the simulated reactor helium supply systems and testing equipment. The progress in the screening test program is descibed; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750/sup 0/, 850/sup 0/, 950/sup 0/ and 1050/sup 0/C (1382/sup 0/, 1562/sup 0/, 1742/sup 0/, and 1922/sup 0/F) in controlled-purity helium. The status of creep-rupture in controlled-purity helium and air and fatigue testing in the controlled-purity helium in the intensive screening test program is discussed. The results of metallographic studies of screening alloys exposed in controlled-purity helium for 3000 hours at 750/sup 0/C and 5500 hours at 950/sup 0/C, 3000 hours at 1050/sup 0/C and 6000 hours at 1050/sup 0/C and for weldments exposed in controlled-purity helium for 6000 hours at 750/sup 0/C and 6000 hours at 1050/sup 0/C are presented and discussed.

  1. Sustainability and the Fixed Bed Nuclear Reactor (FBNR

    Directory of Open Access Journals (Sweden)

    Farhang Sefidvash

    2012-08-01

    Full Text Available Sustainability as a multifaceted and holistic concept is analyzed. Sustainability involves human relationship with elements such as natural environment, economy, power, governance, education and technology with the ultimate purpose of carrying forward an ever-advancing civilization. The Fixed Bed Nuclear Reactor (FBNR is an innovative, small, simple in design, inherently safe, non-proliferating, and environmentally friendly concept that its deployment can generate energy in a sustainable manner contributing to the prosperity of humanity. The development of FBNR will provide electricity as well as desalinated water through a simple but advanced technology for the developing, as well as developed countries. FBNR is environmentally friendly due to its inherent safety and the convenience of using its spent fuel as the source of radiation for irradiation purposes in agriculture, industry, and medicine. Politically, if a ping pong game brought peace between China and USA, a program of development of FBNR supported by the peace loving international community can become a more mature means to bring peace among certain apparently hostile nations who crave sustainable energy, desalinated water and simple advanced technology.

  2. FY2001 Final Report Laboratory Directed Research and Development (LDRD) on Advanced Nuclear Fuel Design in the Future Nuclear Energy Market

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, D.; Choi, J.-S.; DiSabatino, A.; Wirth, B.

    2001-09-30

    This study is to research the maturity of advanced nuclear fuel and cladding technology and to explore the suitability of existing technology for addressing the emerging requirements for Generation IV reactors and emerging thermal/fast spectrum reactors, while simultaneously addressing nuclear waste management, and proliferation resistance concerns.

  3. Physics of nuclear reactors; La physique des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Marguet, S. [Ecole Nationale Superieure de Risques Industriels de Bourges, 18 (France); Institut de Transfert de Technologie d' EDF, 92 - Clamart (France)

    2011-07-01

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  4. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M

    2014-01-01

    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  5. Nuclear data and reactor physics activities in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Liem, P.H. [National Atomic Energy Agency, Tangerang (Indonesia). Center for Multipurpose Reactor

    1998-03-01

    The nuclear data and reactor physics activities in Indonesia, especially, in the National Atomic Energy Agency are presented. In the nuclear data field, the Agency is now taking the position of a user of the main nuclear data libraries such as JENDL and ENDF/B. These nuclear data libraries become the main sources for producing problem dependent cross section sets that are needed by cell calculation codes or transport codes for design, analysis and safety evaluation of research reactors. In the reactor physics field, besides utilising the existing core analysis codes obtained from bilateral and international co-operation, the Agency is putting much effort to self-develop Batan`s codes for reactor physics calculations, in particular, for research reactor and high temperature reactor design, analysis and fuel management. Under the collaboration with JAERI, Monte Carlo criticality calculations on the first criticality of RSG GAS (MPR-30) first core were done using JAERI continuous energy, vectorized Monte Carlo code, MVP, with JENDL-3.1 and JENDL-3.2 nuclear data libraries. The results were then compared with the experiment data collected during the commissioning phase. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k{sub eff} values with excellent agreement with experiment data, however, systematically, JENDL-3.2 library showed slightly higher k{sub eff} values than JENDL-3.1 library. (author)

  6. Nuclear Technology Series. Course 8: Reactor Safety.

    Science.gov (United States)

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutians in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  7. Monitoring Akkuyu Nuclear Reactor Using Anti-Neutrino Flux Measurement

    CERN Document Server

    Ozturk, Sertac; Ozcan, V Erkcan; Unel, Gokhan

    2016-01-01

    We present a simulation based study for monitoring Akkuyu Nuclear Power Plant's activity using anti-neutrino flux originating from the reactor core. A water Cherenkov detector has been designed and optimization studies have been performed using Geant4 simulation toolkit. A first study for the design of a monitoring detector facility for Akkuyu Nuclear Power Plant has been discussed in this paper.

  8. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  9. Economics and utilization of thorium in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    Information on thorium utilization in power reactors is presented concerning the potential demand for nuclear power, the potential supply for nuclear power, economic performance of thorium under different recycle policies, ease of commercialization of the economically preferred cases, policy options to overcome institutional barriers, and policy options to overcome technological and regulatory barriers.

  10. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  11. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  12. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  13. Spent nuclear fuel discharges from US reactors 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  14. Advanced ceramic materials for next-generation nuclear applications

    Science.gov (United States)

    Marra, John

    2011-10-01

    The nuclear industry is at the eye of a 'perfect storm' with fuel oil and natural gas prices near record highs, worldwide energy demands increasing at an alarming rate, and increased concerns about greenhouse gas (GHG) emissions that have caused many to look negatively at long-term use of fossil fuels. This convergence of factors has led to a growing interest in revitalization of the nuclear power industry within the United States and across the globe. Many are surprised to learn that nuclear power provides approximately 20% of the electrical power in the US and approximately 16% of the world-wide electric power. With the above factors in mind, world-wide over 130 new reactor projects are being considered with approximately 25 new permit applications in the US. Materials have long played a very important role in the nuclear industry with applications throughout the entire fuel cycle; from fuel fabrication to waste stabilization. As the international community begins to look at advanced reactor systems and fuel cycles that minimize waste and increase proliferation resistance, materials will play an even larger role. Many of the advanced reactor concepts being evaluated operate at high-temperature requiring the use of durable, heat-resistant materials. Advanced metallic and ceramic fuels are being investigated for a variety of Generation IV reactor concepts. These include the traditional TRISO-coated particles, advanced alloy fuels for 'deep-burn' applications, as well as advanced inert-matrix fuels. In order to minimize wastes and legacy materials, a number of fuel reprocessing operations are being investigated. Advanced materials continue to provide a vital contribution in 'closing the fuel cycle' by stabilization of associated low-level and high-level wastes in highly durable cements, ceramics, and glasses. Beyond this fission energy application, fusion energy will demand advanced materials capable of withstanding the extreme environments of high

  15. Reactor assessments of advanced bumpy torus configurations

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1983-01-01

    Recently, several configurational approaches and concept improvement schemes were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These configurations include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator-snakey torus). Preliminary evaluations of reactor implications of each of these configurations have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties. Results indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past.

  16. Nuclear data uncertainty propagation for a lead-cooled fast reactor: Combining TMC with criticality benchmarks for improved accuracy

    OpenAIRE

    Alhassan, Erwin

    2014-01-01

    For the successful deployment of advanced nuclear systems and for optimization of current reactor designs, high quality and accurate nuclear data are required. Before nuclear data can be used in applications, they are first evaluated, benchmarked against integral experiments and then converted into formats usable for applications. The evaluation process in the past was usually done by using differential experimental data which was then complimented with nuclear model calculations. This trend ...

  17. Nuclear data uncertainty quantification and data assimilation for a lead-cooled fast reactor : Using integral experiments for improved accuracy

    OpenAIRE

    Alhassan, Erwin

    2015-01-01

    For the successful deployment of advanced nuclear systems and optimization of current reactor designs, high quality nuclear data are required. Before nuclear data can be used in applications they must first be evaluated, tested and validated against a set of integral experiments, and then converted into formats usable for applications. The evaluation process in the past was usually done by using differential experimental data which was then complemented with nuclear model calculations. This t...

  18. SNIF: A Futuristic Neutrino Probe for Undeclared Nuclear Fission Reactors

    CERN Document Server

    Lasserre, Thierry; Mention, Guillaume; Reboulleau, Romain; Cribier, Michel; Letourneau, Alain; Lhuillier, David

    2010-01-01

    Today reactor neutrino experiments are at the cutting edge of fundamental research in particle physics. Understanding the neutrino is far from complete, but thanks to the impressive progress in this field over the last 15 years, a few research groups are seriously considering that neutrinos could be useful for society. The International Atomic Energy Agency (IAEA) works with its Member States to promote safe, secure and peaceful nuclear technologies. In a context of international tension and nuclear renaissance, neutrino detectors could help IAEA to enforce the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). In this article we discuss a futuristic neutrino application to detect and localize an undeclared nuclear reactor from across borders. The SNIF (Secret Neutrino Interactions Finder) concept proposes to use a few hundred thousand tons neutrino detectors to unveil clandestine fission reactors. Beyond previous studies we provide estimates of all known background sources as a function of the detecto...

  19. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  20. Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

    2015-10-15

    In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

  1. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  2. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  3. Multiscale Methods for Nuclear Reactor Analysis

    Science.gov (United States)

    Collins, Benjamin S.

    The ability to accurately predict local pin powers in nuclear reactors is necessary to understand the mechanisms that cause fuel pin failure during steady state and transient operation. In the research presented here, methods are developed to improve the local solution using high order methods with boundary conditions from a low order global solution. Several different core configurations were tested to determine the improvement in the local pin powers compared to the standard techniques, that use diffusion theory and pin power reconstruction (PPR). Two different multiscale methods were developed and analyzed; the post-refinement multiscale method and the embedded multiscale method. The post-refinement multiscale methods use the global solution to determine boundary conditions for the local solution. The local solution is solved using either a fixed boundary source or an albedo boundary condition; this solution is "post-refinement" and thus has no impact on the global solution. The embedded multiscale method allows the local solver to change the global solution to provide an improved global and local solution. The post-refinement multiscale method is assessed using three core designs. When the local solution has more energy groups, the fixed source method has some difficulties near the interface: however the albedo method works well for all cases. In order to remedy the issue with boundary condition errors for the fixed source method, a buffer region is used to act as a filter, which decreases the sensitivity of the solution to the boundary condition. Both the albedo and fixed source methods benefit from the use of a buffer region. Unlike the post-refinement method, the embedded multiscale method alters the global solution. The ability to change the global solution allows for refinement in areas where the errors in the few group nodal diffusion are typically large. The embedded method is shown to improve the global solution when it is applied to a MOX/LEU assembly

  4. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  5. Composite Materials under Extreme Radiation and Temperature Environments of the Next Generation Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simos, N.

    2011-05-01

    In the nuclear energy renaissance, driven by fission reactor concepts utilizing very high temperatures and fast neutron spectra, materials with enhanced performance that exceeds are expected to play a central role. With the operating temperatures of the Generation III reactors bringing the classical reactor materials close to their performance limits there is an urgent need to develop and qualify new alloys and composites. Efforts have been focused on the intricate relations and the high demands placed on materials at the anticipated extreme states within the next generation fusion and fission reactors which combine high radiation fluxes, elevated temperatures and aggressive environments. While nuclear reactors have been in operation for several decades, the structural materials associated with the next generation options need to endure much higher temperatures (1200 C), higher neutron doses (tens of displacements per atom, dpa), and extremely corrosive environments, which are beyond the experience on materials accumulated to-date. The most important consideration is the performance and reliability of structural materials for both in-core and out-of-core functions. While there exists a great body of nuclear materials research and operating experience/performance from fission reactors where epithermal and thermal neutrons interact with materials and alter their physio-mechanical properties, a process that is well understood by now, there are no operating or even experimental facilities that will facilitate the extreme conditions of flux and temperature anticipated and thus provide insights into the behaviour of these well understood materials. Materials, however, still need to be developed and their interaction and damage potential or lifetime to be quantified for the next generation nuclear energy. Based on material development advances, composites, and in particular ceramic composites, seem to inherently possess properties suitable for key functions within the

  6. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  7. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    Science.gov (United States)

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described.

  8. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  9. Evaluation of Enhanced Risk Monitors for Use on Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep; Veeramany, Arun; Bonebrake, Christopher A.; Ivans, William J.; Coles, Garill A.; Hirt, Evelyn H.

    2016-09-26

    This study provides an overview of the methodology for integrating time-dependent failure probabilities into nuclear power reactor risk monitors. This prototypic enhanced risk monitor (ERM) methodology was evaluated using a hypothetical probabilistic risk assessment (PRA) model, generated using a simplified design of a liquid-metal-cooled advanced reactor (AR). Component failure data from industry compilation of failures of components similar to those in the simplified AR model were used to initialize the PRA model. Core damage frequency (CDF) over time were computed and analyzed. In addition, a study on alternative risk metrics for ARs was conducted. Risk metrics that quantify the normalized cost of repairs, replacements, or other operations and management (O&M) actions were defined and used, along with an economic model, to compute the likely economic risk of future actions such as deferred maintenance based on the anticipated change in CDF due to current component condition and future anticipated degradation. Such integration of conventional-risk metrics with alternate-risk metrics provides a convenient mechanism for assessing the impact of O&M decisions on safety and economics of the plant. It is expected that, when integrated with supervisory control algorithms, such integrated-risk monitors will provide a mechanism for real-time control decision-making that ensure safety margins are maintained while operating the plant in an economically viable manner.

  10. Primary loop simulation of the SP-100 space nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F., E-mail: eduardo@ieav.cta.b, E-mail: fbraz@ieav.cta.b, E-mail: guimarae@ieav.cta.b [Instituto de Estudos Avancados (IEAv/DCTA) Sao Jose dos Campos, SP (Brazil)

    2011-07-01

    Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

  11. Fractional calculus with applications for nuclear reactor dynamics

    CERN Document Server

    Ray, Santanu Saha

    2015-01-01

    Introduces Novel Applications for Solving Neutron Transport EquationsWhile deemed nonessential in the past, fractional calculus is now gaining momentum in the science and engineering community. Various disciplines have discovered that realistic models of physical phenomenon can be achieved with fractional calculus and are using them in numerous ways. Since fractional calculus represents a reactor more closely than classical integer order calculus, Fractional Calculus with Applications for Nuclear Reactor Dynamics focuses on the application of fractional calculus to describe the physical behavi

  12. Development of a research nuclear reactor simulator using LABVIEW®

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

  13. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Reactor Physics Dept.

    2014-12-15

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  14. China Advanced Research Reactor Project Progress in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    2011, China Advanced Research Reactor (CARR) Project finished the B stage commissioning and resolved the relative technical problems. Meanwhile, the acceptance items and the cold neutron source were carrying out.

  15. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  16. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  17. Reactor assessments of advanced bumpy torus configurations

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1984-02-01

    Recently, several innovative approaches were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator - snakey torus). Preliminary evaluations of reactor implications of each approach have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties deduced from provisional configurations that implement the approach but are not necessarily optimized. Further optimization is needed in all cases to evaluate the full potential of each approach. Results of these studies indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past.

  18. Spectral Structure of Electron Antineutrinos from Nuclear Reactors

    CERN Document Server

    Dwyer, D A

    2014-01-01

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principle calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructure in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of this substructure can constrain nuclear reactor physics. The substructure can be a systematic uncertainty for measurements utilizing the detailed spectral shape.

  19. Spectral structure of electron antineutrinos from nuclear reactors.

    Science.gov (United States)

    Dwyer, D A; Langford, T J

    2015-01-01

    Recent measurements of the positron energy spectrum obtained from inverse beta decay interactions of reactor electron antineutrinos show an excess in the 4 to 6 MeV region relative to current predictions. First-principles calculations of fission and beta decay processes within a typical pressurized water reactor core identify prominent fission daughter isotopes as a possible origin for this excess. These calculations also predict percent-level substructures in the antineutrino spectrum due to Coulomb effects in beta decay. Precise measurement of these substructures can elucidate the nuclear processes occurring within reactors. These substructures can be a systematic issue for measurements utilizing the detailed spectral shape.

  20. Recent Canadian advances in nuclear-based hydrogen production and the thermochemical Cu-Cl cycle

    Energy Technology Data Exchange (ETDEWEB)

    Naterer, G. [Canada Research Chair Professor, University of Ontario Institute of Technology (UOIT), 2000 Simcoe Street, Oshawa, Ontario L1H 7K4 (Canada); Suppiah, S. [Manager, Hydrogen Isotopes Technology Branch, AECL, Chalk River, Ontario K0J 1J0 (Canada); Lewis, M. [Chemist, Chemical Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, Illinois 60439 (United States); Gabriel, K. [Associate Provost, Research, UOIT, 2000 Simcoe Street North, Oshawa, Ontario L1H 7K4 (Canada); Dincer, I.; Rosen, M.A. [Professor of Mechanical Engineering, UOIT, 2000 Simcoe Street North, Oshawa, Ontario L1H 7K4 (Canada); Fowler, M. [Assistant Professor of Chemical Engineering, University of Waterloo, 200 University Avenue, Waterloo, Ontario N2L 3G1 (Canada); Rizvi, G. [Assistant Professor of Mechanical Engineering, UOIT, 2000 Simcoe Street North, Oshawa, Ontario L1H 7K4 (Canada); Easton, E.B. [Assistant Professor of Chemistry, UOIT, 2000 Simcoe Street North, Oshawa, Ontario L1H 7K4 (Canada); Ikeda, B.M.; Pioro, I. [Associate Professor, Faculty of Energy Systems and Nuclear Science, UOIT, 2000 Simcoe St., Oshawa, ON L1H 7K4 (Canada); Kaye, M.H.; Lu, L. [Assistant Professor, Faculty of Energy Systems and Nuclear Science, UOIT, 2000 Simcoe Street, Oshawa, Ontario L1H 7K4 (Canada); Spekkens, P. [Vice President of Science and Technology Development, Ontario Power Generation, 889 Brock Road, Pickering, Ontario (Canada); Tremaine, P. [Professor of Chemistry, University of Guelph, 50 Stone Road East, Guelph, Ontario N1G 2W1 (Canada); Mostaghimi, J. [Canada Research Chair Professor, Mechanical Engineering, University of Toronto, Toronto, Ontario M5S 3E5 (Canada); Avsec, J. [Assistant Professor, Faculty of Energy Technology, Univ. of Maribor, Hocevarjev trg 1, 8270 Krsko (Slovenia); Jiang, J. [Professor and NSERC/UNENE Senior Industrial Research Chair, Electrical and Computer Engineering, Univ. of Western Ontario, London, Ontario N6A 5B9 (Canada)

    2009-04-15

    This paper presents recent Canadian advances in nuclear-based production of hydrogen by electrolysis and the thermochemical copper-chlorine (Cu-Cl) cycle. This includes individual process and reactor developments within the Cu-Cl cycle, thermochemical properties, advanced materials, controls, safety, reliability, economic analysis of electrolysis at off-peak hours, and integrating hydrogen plants with Canada's nuclear power plants. These enabling technologies are being developed by a Canadian consortium, as part of the Generation IV International Forum (GIF) for hydrogen production from the next generation of nuclear reactors. (author)

  1. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  2. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  3. Exploring Stochastic Sampling in Nuclear Data Uncertainties Assessment for Reactor Physics Applications and Validation Studies

    Directory of Open Access Journals (Sweden)

    Alexander Vasiliev

    2016-12-01

    Full Text Available The quantification of uncertainties of various calculation results, caused by the uncertainties associated with the input nuclear data, is a common task in nuclear reactor physics applications. Modern computation resources and improved knowledge on nuclear data allow nowadays to significantly advance the capabilities for practical investigations. Stochastic sampling is the method which has received recently a high momentum for its use and exploration in the domain of reactor design and safety analysis. An application of a stochastic sampling based tool towards nuclear reactor dosimetry studies is considered in the given paper with certain exemplary test evaluations. The stochastic sampling not only allows the input nuclear data uncertainties propagation through the calculations, but also an associated correlation analysis performance with no additional computation costs and for any parameters of interest can be done. Thus, an example of assessment of the Pearson correlation coefficients for several models, used in practical validation studies, is shown here. As a next step, the analysis of the obtained information is proposed for discussion, with focus on the systems similarities assessment. The benefits of the employed method and tools with respect to practical reactor dosimetry studies are consequently outlined.

  4. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1962-01-01

    Advances in Nuclear Science and Technology, Volume 1 provides an authoritative, complete, coherent, and critical review of the nuclear industry. This book covers a variety of topics, including nuclear power stations, graft polymerization, diffusion in uranium alloys, and conventional power plants.Organized into seven chapters, this volume begins with an overview of the three stages of the operation of a power plant, either nuclear or conventionally fueled. This text then examines the major problems that face the successful development of commercial nuclear power plants. Other chapters consider

  5. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    Science.gov (United States)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  6. Nuclear Thermal Propulsion for Advanced Space Exploration

    Science.gov (United States)

    Houts, M. G.; Borowski, S. K.; George, J. A.; Kim, T.; Emrich, W. J.; Hickman, R. R.; Broadway, J. W.; Gerrish, H. P.; Adams, R. B.

    2012-01-01

    The fundamental capability of Nuclear Thermal Propulsion (NTP) is game changing for space exploration. A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on NTP could provide high thrust at a specific impulse above 900 s, roughly double that of state of the art chemical engines. Characteristics of fission and NTP indicate that useful first generation systems will provide a foundation for future systems with extremely high performance. The role of the NCPS in the development of advanced nuclear propulsion systems could be analogous to the role of the DC-3 in the development of advanced aviation. Progress made under the NCPS project could help enable both advanced NTP and advanced Nuclear Electric Propulsion (NEP).

  7. Production capabilities in US nuclear reactors for medical radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. [Oak Ridge National Lab., TN (United States); Schenter, R.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-11-01

    The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

  8. Global risk of radioactive fallout after major nuclear reactor accidents

    Science.gov (United States)

    Lelieveld, J.; Kunkel, D.; Lawrence, M. G.

    2012-05-01

    Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7), using particulate 137Cs and gaseous 131I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  9. Global risk of radioactive fallout after major nuclear reactor accidents

    Directory of Open Access Journals (Sweden)

    J. Lelieveld

    2012-05-01

    Full Text Available Major reactor accidents of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the cumulative, global risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents (the most severe ones on the International Nuclear Event Scale, INES 7, using particulate 137Cs and gaseous 131I as proxies for the fallout. Our results indicate that previously the occurrence of INES 7 major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a major reactor accident of any nuclear power plant worldwide, more than 90% of emitted 137Cs would be transported beyond 50 km and about 50% beyond 1000 km distance before being deposited. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human exposure due to deposition are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in West Europe and South Asia, where a major reactor accident can subject around 30 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  10. Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: alpsordo@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2012-10-15

    The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)

  11. Handbook of nuclear engineering: vol 1: nuclear engineering fundamentals; vol 2: reactor design; vol 3: reactor analysis; vol 4: reactors of waste disposal and safeguards

    CERN Document Server

    2013-01-01

    The Handbook of Nuclear Engineering is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all academic levels, this five volume set provides the latest findings in nuclear data and experimental techniques, reactor physics, kinetics, dynamics and control. Readers will also find a detailed description of data assimilation, model validation and calibration, sensitivity and uncertainty analysis, fuel management and cycles, nuclear reactor types and radiation shielding. A discussion of radioactive waste disposal, safeguards and non-proliferation, and fuel processing with partitioning and transmutation is also included. As nuclear technology becomes an important resource of non-polluting sustainable energy in the future, The Handbook of Nuclear Engineering is an excellent reference for practicing engineers, researchers and professionals.

  12. Modelling of advanced structural materials for GEN IV reactors

    Science.gov (United States)

    Samaras, M.; Hoffelner, W.; Victoria, M.

    2007-09-01

    The choice of suitable materials and the assessment of long-term materials damage are key issues that need to be addressed for the safe and reliable performance of nuclear power plants. Operating conditions such as high temperatures, irradiation and a corrosive environment degrade materials properties, posing the risk of very expensive or even catastrophic plant damage. Materials scientists are faced with the scientific challenge to determine the long-term damage evolution of materials under service exposure in advanced plants. A higher confidence in life-time assessments of these materials requires an understanding of the related physical phenomena on a range of scales from the microscopic level of single defect damage effects all the way up to macroscopic effects. To overcome lengthy and expensive trial-and-error experiments, the multiscale modelling of materials behaviour is a promising tool, bringing new insights into the fundamental understanding of basic mechanisms. This paper presents the multiscale modelling methodology which is taking root internationally to address the issues of advanced structural materials for Gen IV reactors.

  13. Nuclear reactor (1960); Reacteurs nucleaires (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Maillard, M.L. [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Leo, M.B. [Electricite de France (EDF), 75 - Paris (France)

    1960-07-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [French] Les premiers reacteurs industriels plutonigenes francais G1 - G2 - G3 du Centre de Marcoule comportent une installation de recuperation d'energie. La production d'electricite de G1 ne compense pas l'energie depensee par ailleurs pour le fonctionnement de l'ensemble, par contre, G2 et G3 doivent fournir chacun une puissance de 25 a 30 MW au reseau national d'Electricite de France. Cette puissance est modeste, mais l'experience acquise grace a ces reacteurs est tres grande et c'est grace a elle qu'il nous sera possible de mettre en exploitation les reacteurs energetiques EDF1 - EDF2 - EDF3. Le memoire decrit comment, avant tout demarrage du reacteur, les essais effectues, en particulier ceux concernant l'installation de recuperation d'energie et le caisson, ont permis d'abreger la phase de montee en puissance. (auteur)

  14. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  15. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  16. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko

    2012-09-01

    Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

  17. Neutron Capture and the Antineutrino Yield from Nuclear Reactors

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ˜0.9 % of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  18. Detectability prediction for a thermoacoustic sensor in the breazeale nuclear reactor pool

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James [Idaho National Laboratory, Idaho Falls, ID (United States); Hrisko, Joshua [Idaho National Laboratory, Idaho Falls, ID (United States); Garrett, Steven [Idaho National Laboratory, Idaho Falls, ID (United States)

    2016-03-01

    Laboratory experiments have suggested that thermoacoustic engines can be in- corporated within nuclear fuel rods. Such engines would radiate sounds that could be used to measure and acoustically-telemeter information about the op- eration of the nuclear reactor (e.g., coolant temperature or uxes of neutrons or other energetic particles) or the physical condition of the nuclear fuel itself (e.g., changes in temperature, evolved gases) that are encoded as the frequency and/or amplitude of the radiated sound [IEEE Measurement and Instrumen- tation 16(3), 18-25 (2013)]. For such acoustic information to be detectable, it is important to characterize the vibroacoustical environments within reactors. Measurements will be presented of the background noise spectra (with and with- out coolant pumps) and reverberation times within the 70,000 gallon pool that cools and shields the fuel in the 1 MW research reactor on Penn State's campus using two hydrophones, a piezoelectric projector, and an accelerometer. Sev- eral signal-processing techniques will be demonstrated to enhance the measured results. Background vibrational measurement were also taken at the 250 MW Advanced Test Reactor, located at the Idaho National Laboratory, using ac- celerometers mounted outside the reactor's pressure vessel and on plumbing will also be presented. The detectability predictions made in the thesis were validated in September 2015 using a nuclear ssion-heated thermoacoustic sensor that was placed in the core of the Breazeale Nuclear Reactor on Penn State's campus. Some features of the thermoacoustic device used in that experiment will also be revealed. [Work supported by the U.S. Department of Energy.

  19. Piezoelectric material for use in a nuclear reactor core

    Science.gov (United States)

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-01

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d33 was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d33 for many as-grown samples.

  20. Alloying of steel and graphite by hydrogen in nuclear reactor

    Science.gov (United States)

    Krasikov, E.

    2017-02-01

    In traditional power engineering hydrogen may be one of the first primary source of equipment damage. This problem has high actuality for both nuclear and thermonuclear power engineering. Study of radiation-hydrogen embrittlement of the steel raises the question concerning the unknown source of hydrogen in reactors. Later unexpectedly high hydrogen concentrations were detected in irradiated graphite. It is necessary to look for this source of hydrogen especially because hydrogen flakes were detected in reactor vessels of Belgian NPPs. As a possible initial hypothesis about the enigmatical source of hydrogen one can propose protons generation during beta-decay of free neutrons поскольку inasmuch as protons detected by researches at nuclear reactors as witness of beta-decay of free neutrons.

  1. Optimizing Nuclear Reactor Operation Using Soft Computing Techniques

    NARCIS (Netherlands)

    Entzinger, J.O.; Ruan, D.; Kahraman, Cengiz

    2006-01-01

    The strict safety regulations for nuclear reactor control make it di±cult to implement new control techniques such as fuzzy logic control (FLC). FLC however, can provide very desirable advantages over classical control, like robustness, adaptation and the capability to include human experience into

  2. Global risk of radioactive fallout after nuclear reactor accidents

    Science.gov (United States)

    Kunkel, D.; Lelieveld, J.; Lawrence, M. G.

    2012-04-01

    Reactor core meltdowns of nuclear power plants are rare, yet the consequences are catastrophic. But what is meant by "rare"? And what can be learned from the Chernobyl and Fukushima incidents? Here we assess the risk of exposure to radioactivity due to atmospheric dispersion of gases and particles following severe nuclear accidents, using particulate 137Cs and gaseous 131I as proxies for the fallout. It appears that previously the occurrence of major accidents and the risks of radioactive contamination have been underestimated. Using a global model of the atmosphere we compute that on average, in the event of a core melt of any nuclear power plant worldwide, more than 90 % of emitted 137Cs would be transported beyond 50 km and about 50 % beyond 1000 km distance. This corroborates that such accidents have large-scale and trans-boundary impacts. Although the emission strengths and atmospheric removal processes of 137Cs and 131I are quite different, the radioactive contamination patterns over land and the human deposition exposure are computed to be similar. High human exposure risks occur around reactors in densely populated regions, notably in southern Asia where a core melt can subject 55 million people to radioactive contamination. The recent decision by Germany to phase out its nuclear reactors will reduce the national risk, though a large risk will still remain from the reactors in neighbouring countries.

  3. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  4. Use of hafnium in control bars of nuclear reactors; Uso de hafnio en barras de control de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin-mx

    2003-07-01

    Recently the use of hafnium as neutron absorber material in nuclear reactors has been reason of investigation by virtue of that this material has nuclear properties as to the neutrons absorption and structural that can prolong the useful life of the control mechanisms of the nuclear reactors. In this work some of those more significant hafnium properties are presented like nuclear material. Also there are presented calculations carried out with the HELIOS code for fuel cells of uranium oxide and of uranium and plutonium mixed oxides under controlled conditions with conventional bars of boron carbide and also with similar bars to which are substituted the absorbent material by metallic hafnium, the results are presented in this work. (Author)

  5. Advances in nuclear science and technology

    CERN Document Server

    Greebler, Paul

    1966-01-01

    Advances in Nuclear Science and Technology, Volume 3 provides an authoritative, complete, coherent, and critical review of the nuclear industry. This book presents the advances in the atomic energy field.Organized into six chapters, this volume begins with an overview of the use of pulsed neutron sources for the determination of the thermalization and diffusion properties of moderating as well as multiplying media. This text then examines the effect of nuclear radiation on electronic circuitry and its components. Other chapters consider radiation effects in various inorganic solids, with empha

  6. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    Energy Technology Data Exchange (ETDEWEB)

    Bohachek, Randolph Charles [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  7. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    Energy Technology Data Exchange (ETDEWEB)

    Loika, E.F.

    1994-09-22

    This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate.

  8. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  9. Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

    Energy Technology Data Exchange (ETDEWEB)

    Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

    2015-06-15

    Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

  10. 77 FR 39521 - Application for a License To Export Nuclear Reactor Major Components and Equipment

    Science.gov (United States)

    2012-07-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Application for a License To Export Nuclear Reactor Major Components and Equipment Pursuant to 10... Reactor internals, Components and For use in Braka nuclear power Company LLC reactor coolant equipment...

  11. Advanced reactor material research requirements; Necesidades de investigacion en materiales para reactores avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Greene, C. A.; Muscara, J.; Srinivasan, M.

    2003-07-01

    The metal and graphite components used in high temperature gas-cooled reactors (HTGR) may suffer physical-chemical alterations, irradiation damage and mechanical alterations. Their failure may call the security of these reactors into question by affecting the integrity of the pressure control system, core geometry or its cooling, among other aspects. This article analyses the work currently being done in the matter by the US Nuclear Regulatory Commission. (Author)

  12. Reference site selection report for the advanced liquid metal reactor at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Sivill, R.L.

    1990-03-01

    This Reference Site Selection Report was prepared by EG G, Idaho Inc., for General Electric (GE) to provide information for use by the Department of Energy (DOE) in selecting a Safety Test Site for an Advanced Liquid Metal Reactor. Similar Evaluation studies are planned to be conducted at other potential DOE sites. The Power Reactor Innovative Small Module (PRISM) Concept was developed for ALMR by GE. A ALMR Safety Test is planned to be performed on a DOE site to demonstrate features and meet Nuclear Regulatory Commission Requirements. This study considered possible locations at the Idaho National Engineering Laboratory that met the ALMR Prototype Site Selection Methodology and Criteria. Four sites were identified, after further evaluation one site was eliminated. Each of the remaining three sites satisfied the criteria and was graded. The results were relatively close. Thus concluding that the Idaho National Engineering Laboratory is a suitable location for an Advanced Liquid Metal Reactor Safety Test. 23 refs., 13 figs., 9 tabs.

  13. Methodology for the integral comparison of nuclear reactors: selecting a reactor for Mexico; Metodologia para la comparacion integral de reactores nucleares: seleccion de un reactor para Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: ricarera@yahoo.com.mx

    2006-07-01

    In this work it was built a methodology to compare nuclear reactors of third generation that can be contemplated for future electric planning in Mexico. This methodology understands the selection of the reactors to evaluate eliminating the reactors that still are not sufficiently mature at this time of the study. A general description of each reactor together with their main ones characteristic is made. It was carried out a study for to select the group of parameters that can serve as evaluation indicators, which are the characteristics of the reactors with specific values for each considered technology, and it was elaborated an evaluation matrix indicators including the reactors in the columns and those indicators in the lines. Since that none reactor is the best in all the indicators were necessary to use a methodology for multi criteria taking decisions that we are presented. It was used the 'Fuzzy Logic' technique, the which is based in those denominated diffuse groups and in a system of diffuse inference based on heuristic rules in the way 'If Then consequence> ', where the linguistic values of the condition and of the consequence is defined by diffuse groups, it is as well as the rules always they transform a diffuse group into another. Later on they combine all the diffuse outputs to create a single output and an inverse transformation is made that it transfers the output from the diffuse domain to the real one. They were carried out two studies one with the entirety of the indicators and another in which the indicators were classified in three approaches: the first one refers to indicators related with the planning of the plants inside the context of the general electric sector, the second approach includes indicators related with the characteristics of the fuel and the third it considers indicators related with the acting of the plant in safety and environmental impact. This second study allowed us to know the qualities of

  14. Advanced Reactors Transition program fiscal year 1998 multi-year work plan

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D.A.

    1997-09-25

    The mission of the Advanced Reactors Transition program is two-fold. First, the program is to maintain the Fast Flux Test Facility (FFTF) and the Fuels and Materials Examination Facility (FMEF) in Standby to support a possible future role in the tritium production strategy. Secondly, the program is to continue deactivation activities which do not conflict with the Standby directive. On-going deactivation activities include the processing of non-usable, irradiated, FFTF components for storage or disposal; deactivation of Nuclear Energy legacy test facilities; and deactivation of the Plutonium Recycle Test Reactor (PRTR) facility, 309 Building.

  15. Neutron capture and the antineutrino yield from nuclear reactors

    CERN Document Server

    Huber, Patrick

    2015-01-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low-energies below 3.2MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach 0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the non-equilibrium correction...

  16. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

  17. Nuclear reactor melt arrest and coolability device

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, Theo G.; Dinh, Nam Truc; Wachowiak, Richard M.

    2016-06-14

    Example embodiments provide a Basemat-Internal Melt Arrest and Coolability device (BiMAC) that offers improved spatial and mechanical characteristics for use in damage prevention and risk mitigation in accident scenarios. Example embodiments may include a BiMAC having an inclination of less than 10-degrees from the basemat floor and/or coolant channels of less than 4 inches in diameter, while maintaining minimum safety margins required by the Nuclear Regulatory Commission.

  18. Parallelization and automatic data distribution for nuclear reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Liebrock, L.M. [Liebrock-Hicks Research, Calumet, MI (United States)

    1997-07-01

    Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed.

  19. ORNL R and D on advanced small and medium power reactors: Selected topics

    Energy Technology Data Exchange (ETDEWEB)

    White, J.D.; Trauger, D.B.

    1988-01-01

    From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, and assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable RandD would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current RandD efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described. 13 refs., 1 fig.

  20. Nuclear reactor for breeding U.sup.233

    Science.gov (United States)

    Bohanan, Charles S.; Jones, David H.; Raab, Jr., Harry F.; Radkowsky, Alvin

    1976-01-01

    A light-water-cooled nuclear reactor capable of breeding U.sup.233 for use in a light-water breeder reactor includes physically separated regions containing U.sup.235 fissile material and U.sup.238 fertile material and Th.sup.232 fertile material and Pu.sup.239 fissile material, if available. Preferably the U.sup.235 fissile material and U.sup.238 fertile material are contained in longitudinally movable seed regions and the Pu.sup.239 fissile material and Th.sup.232 fertile material are contained in blanket regions surrounding the seed regions.

  1. iDREAM: an industrial detector for nuclear reactor monitoring

    Science.gov (United States)

    Gribov, I. V.; Gromov, M. B.; Lukjanchenko, G. A.; Novikova, G. J.; Obinyakov, B. A.; Oralbaev, A. Y.; Skorokhvatov, M. D.; Sukhotin, S. V.; Chepurnov, A. S.; Etenko, A. V.

    2016-02-01

    Prototype of industrial reactor antineutrino detector iDREAM is dedicated for an experiment to demonstrate the possibility of remote monitoring of PWR reactor operational modes by neutrino method in real-time in order to avoid undeclared exposure modes for nuclear fuel and unauthorized removal of isotopes. The prototype detector was started up in 2014. To test the detector elements and components of electronics distilled water has been used as a target, which enables the use of Cerenkov radiation from cosmic muons as a physical signal. Also parallel measuring of the long-term stability has been doing for samples of liquid organic scintillator doped with gadolinium and synthesized by different methods

  2. Transient behavior of a nuclear reactor coupled to an accelerator

    Science.gov (United States)

    Sadineni, Suresh Babu

    Accelerator Driven Systems (ADS) present one of the most viable solutions for transmutation and effective utilization of nuclear fuel. Spent fuel from reactors will be partitioned to separate plutonium and other minor actinides to be transmuted in the ADS. Without the ADS, minor actinides must be stored at a geologic repository for long periods of time. One problem with ADS is understanding the control issues that arise when coupling an accelerator to a reactor. "ADSTRANS" was developed to predict the transient behavior of a nuclear reactor coupled to an accelerator. It was based on MCNPX, a radiation transport code developed at the LANL, and upon a numerical model of the neutron transport equation. MCNPX was used to generate the neutron "source" term that occurs when the accelerator is fired. ADSTRANS coupled MCNPX to a separate finite difference code that solved the transient neutron transport equation. A cylindrical axisymmetric reactor with steel shielding was considered for this analysis. Multiple neutron energy groups, neutron precursor groups and neutron poisons were considered. ENDF/B cross-section data obtained through MCNPX was also employed. The reactor was assumed to be isothermal and near zero power level. Unique features of this code are: (1) it predicts the neutron behavior of an ADS for different reactor geometry, material concentration, both electron and proton particle accelerators, and target material, (2) it develops input files for MCNPX to simulate neutron production, runs MCNPX, and retrieves information from the MCNPX output files. Neutron production predicted by MCNPX for a 20 MeV electron accelerator and lead target was compared with experimental data from the Idaho Accelerator Center and found to be in good agreement. The spatial neutron flux distribution and transient neutron flux in the reactor as predicted by the code were compared with analytical solutions and found to be in good agreement. Fuel burnup and poison buildup were also as

  3. Basic Model of a Control Assembly Drop in Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Radek BULÍN

    2013-06-01

    Full Text Available This paper is focused on the modelling and dynamic analysis of a nonlinear system representing a control assembly of the VVER 440/V213 nuclear reactor. A simple rigid body model intended for basic dynamic analyses is introduced. It contains the influences of the pressurized water and mainly the eects of possible control assembly contacts with guiding tubes inside the reactor. Another approach based on a complex multibody model is further described and the suitability of both modelling approaches is discussed.

  4. Thermohydraulic and nuclear modeling of natural fission reactors

    Science.gov (United States)

    Viggato, Jason Charles

    Experimental verification of proposed nuclear waste storage schemes in geologic repositories is not possible, however, a natural analog exists in the form of ancient natural reactors that existed in uranium-rich ores. Two billion years ago, the enrichment of natural uranium was high enough to allow a sustained chain reaction in the presence of water as a moderator. Several natural reactors occurred in Gabon, Africa and were discovered in the early 1970's. These reactors operated at low power levels for hundreds of thousands of years. Heated water generated from the reactors also leached uranium from the surrounding rock strata and deposited it in the reactor cores. This increased the concentration of uranium in the core over time and served to "refuel" the reactor. This has strong implications in the design of modern geologic repositories for spent nuclear fuel. The possibility of accidental fission events in man-made repositories exists and the geologic evidence from Oklo suggests how those events may progress and enhance local concentrations of uranium. Based on a review of the literature, a comprehensive code was developed to model the thermohydraulic behavior and criticality conditions that may have existed in the Oklo reactor core. A two-dimensional numerical model that incorporates modeling of fluid flow, temperatures, and nuclear fission and subsequent heat generation was developed for the Oklo natural reactors. The operating temperatures ranged from about 456 K to about 721 K. Critical reactions were observed for a wide range of concentrations and porosity values (9 to 30 percent UO2 and 10 to 20 percent porosity). Periodic operation occurred in the computer model prediction with UO2 concentrations of 30 percent in the core and 5 percent in the surrounding material. For saturated conditions and 30 percent porosity, the model predicted temperature transients with a period of about 5 hours. Kuroda predicted 3 to 4 hour durations for temperature transients

  5. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  6. Modeling and Testing of Non-Nuclear, Highpower Simulated Nuclear Thermal Rocket Reactor Elements

    Science.gov (United States)

    Kirk, Daniel R.

    2005-01-01

    When the President offered his new vision for space exploration in January of 2004, he said, "Our third goal is to return to the moon by 2020, as the launching point for missions beyond," and, "With the experience and knowledge gained on the moon, we will then be ready to take the next steps of space exploration: human missions to Mars and to worlds beyond." A human mission to Mars implies the need to move large payloads as rapidly as possible, in an efficient and cost-effective manner. Furthermore, with the scientific advancements possible with Project Prometheus and its Jupiter Icy Moons Orbiter (JIMO), (these use electric propulsion), there is a renewed interest in deep space exploration propulsion systems. According to many mission analyses, nuclear thermal propulsion (NTP), with its relatively high thrust and high specific impulse, is a serious candidate for such missions. Nuclear rockets utilize fission energy to heat a reactor core to very high temperatures. Hydrogen gas flowing through the core then becomes superheated and exits the engine at very high exhaust velocities. The combination of temperature and low molecular weight results in an engine with specific impulses above 900 seconds. This is almost twice the performance of the LOX/LH2 space shuttle engines, and the impact of this performance would be to reduce the trip time of a manned Mars mission from the 2.5 years, possible with chemical engines, to about 12-14 months.

  7. Computer code applicability assessment for the advanced Candu reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wren, D.J.; Langman, V.J.; Popov, N.; Snell, V.G. [Atomic Energy of Canada Ltd (Canada)

    2004-07-01

    AECL Technologies, the 100%-owned US subsidiary of Atomic Energy of Canada Ltd. (AECL), is currently the proponents of a pre-licensing review of the Advanced Candu Reactor (ACR) with the United States Nuclear Regulatory Commission (NRC). A key focus topic for this pre-application review is the NRC acceptance of the computer codes used in the safety analysis of the ACR. These codes have been developed and their predictions compared against experimental results over extended periods of time in Canada. These codes have also undergone formal validation in the 1990's. In support of this formal validation effort AECL has developed, implemented and currently maintains a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper discusses the SQA program used to develop, qualify and maintain the computer codes used in ACR safety analysis, including the current program underway to confirm the applicability of these computer codes for use in ACR safety analyses. (authors)

  8. On-Line NDE for Advanced Reactor Designs

    Science.gov (United States)

    Nakagawa, N.; Inanc, F.; Thompson, R. B.; Junker, W. R.; Ruddy, F. H.; Beatty, J. M.; Arlia, N. G.

    2003-03-01

    This expository paper introduces the concept of on-line sensor methodologies for monitoring the integrity of components in next generation power systems, and explains general benefits of the approach, while describing early conceptual developments of suitable NDE methodologies. The paper first explains the philosophy behind this approach (i.e. the design-for-inspectability concept). Specifically, we describe where and how decades of accumulated knowledge and experience in nuclear power system maintenance are utilized in Generation IV power system designs, as the designs are being actively developed, in order to advance their safety and economy. Second, we explain that Generation IV reactor design features call for the replacement of the current outage-based maintenance by on-line inspection and monitoring. Third, the model-based approach toward design and performance optimization of on-line sensor systems, using electromagnetic, ultrasonic, and radiation detectors, will be explained. Fourth, general types of NDE inspections that are considered amenable to on-line health monitoring will be listed. Fifth, we will describe specific modeling developments to be used for radiography, EMAT UT, and EC detector design studies.

  9. Synergistic smart fuel for in-pile nuclear reactor measurements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

    2013-07-01

    The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

  10. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-10-06

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  11. Optimization Algorithms for Nuclear Reactor Power Control

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Min; Oh, Won Jong; Oh, Seung Jin; Chun, Won Gee; Lee, Yoon Joon [Jeju National University, Jeju (Korea, Republic of)

    2010-10-15

    One of the control techniques that could replace the present conventional PID controllers in nuclear plants is the linear quadratic regulator (LQR) method. The most attractive feature of the LQR method is that it can provide the systematic environments for the control design. However, the LQR approach heavily depends on the selection of cost function and the determination of the suitable weighting matrices of cost function is not an easy task, particularly when the system order is high. The purpose of this paper is to develop an efficient and reliable algorithm that could optimize the weighting matrices of the LQR system

  12. Application of probabilistic risk assessment in nuclear and environmental licensing processes of nuclear reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Mata, Jonatas F.C. da; Vasconcelos, Vanderley de; Mesquita, Amir Z., E-mail: jonatasfmata@yahoo.com.br, E-mail: vasconv@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The nuclear accident at Fukushima Daiichi, occurred in Japan in 2011, brought reflections, worldwide, on the management of nuclear and environmental licensing processes of existing nuclear reactors. One of the key lessons learned in this matter, is that the studies of Probabilistic Safety Assessment and Severe Accidents are becoming essential, even in the early stage of a nuclear development project. In Brazil, Brazilian Nuclear Energy Commission, CNEN, conducts the nuclear licensing. The organism responsible for the environmental licensing is Brazilian Institute of Environment and Renewable Natural Resources, IBAMA. In the scope of the licensing processes of these two institutions, the safety analysis is essentially deterministic, complemented by probabilistic studies. The Probabilistic Safety Assessment (PSA) is the study performed to evaluate the behavior of the nuclear reactor in a sequence of events that may lead to the melting of its core. It includes both probability and consequence estimation of these events, which are called Severe Accidents, allowing to obtain the risk assessment of the plant. Thus, the possible shortcomings in the design of systems are identified, providing basis for safety assessment and improving safety. During the environmental licensing, a Quantitative Risk Analysis (QRA), including probabilistic evaluations, is required in order to support the development of the Risk Analysis Study, the Risk Management Program and the Emergency Plan. This article aims to provide an overview of probabilistic risk assessment methodologies and their applications in nuclear and environmental licensing processes of nuclear reactors in Brazil. (author)

  13. Hiberarchy of Requi rement Analysis on Diversity Actuation System for Advanced Pressurized Water Reactor Nuclear Power Plant%先进压水堆核电厂多样性驱动系统需求分析结构

    Institute of Scientific and Technical Information of China (English)

    刘爱国; 江辉; 孙伟; 田亚杰; 朱学农

    2014-01-01

    核电厂数字化应用需应对软件可能的共因失效所带来的问题。多样性驱动系统(DAS )用于数字化保护系统软件共因失效(CCF)时缓解和抵御核电厂发生设计基准事件的后果。本研究在核电厂纵深防御与多样性原则的指导下,对先进压水堆核电厂的多样性驱动系统需求进行分析研究,阐明了先进压水堆核电厂多样性驱动系统的设计特性、过程、依据、方法及系统结构。%The digital systems applied to nuclear power plants need to address the prob‐lems created by common‐cause failure (CCF) of the software .The diversity actuation system (DAS) can resist and relieve accident consequence for nuclear power plant ,when postulated CCF of reactor protection system and design‐basis event take place at the same time .According to the defense‐in‐depth and diversity principles of nuclear power plant ,the requirement analysis of DAS for pressurized water reactor was researched . The design characteristic property ,design process ,design basis ,design methods and system structure of DAS were systematically expounded .

  14. Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

    Energy Technology Data Exchange (ETDEWEB)

    Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

  15. Strategic Need for Multi-Purpose Thermal Hydraulic Loop for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; Su-Jong Yoon; Gregory K. Housley

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  16. Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  17. Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    OHara J. M.; Higgins, J.; DAgostino, A.

    2012-01-17

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

  18. Simulating experimental investigation on the safety of nuclear heating reactor in loss-of-coolant accidents

    Science.gov (United States)

    Xu, Zhanjie

    1996-12-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. Up to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as follos: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  19. Advanced Reactor Safety Program – Stakeholder Interaction and Feedback

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    In the Spring of 2013, the Idaho National Laboratory (INL) began discussions with industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling and simulation could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

  20. Advanced reactor safety program. Stakeholder interaction and feedback

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H.; Smith, Curtis L.

    2014-08-01

    In the Spring of 2013, we began discussions with our industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require substantial participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

  1. Sub-Critical Nuclear Reactor Based on FFAG-Accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee Seok; Kang, Hung Sik; Lee, Tae Yeon [Pohang Accelerator Laboratory, Pohang (Korea, Republic of)

    2011-10-15

    After the East-Japan earthquake and the subsequent nuclear disaster, the anti-nuclear mood has been wide spread. It is very unfortunate both for nuclear science community and for the future of mankind, which is threatened by two serious challenges, the global warming caused by the greenhouse effect and the shortage of energy cause by the petroleum exhaustion. While the nuclear energy seemed to be the only solution to these problems, it is clear that it has its own problems, one of which broke out so strikingly in Japan. There are also other problems such as the radiotoxic nuclear wastes that survive up to even tens of thousands years and the limited reserves of Uranium. To solve these problems of nuclear fission energy, accelerator-based sub-critical nuclear reactor was once proposed. (Its details will be explained below.) First of all, it is safe in a disaster such as an earthquake, because the deriving accelerator stops immediately by the earthquake. It also minimizes the nuclear waste problem by reducing the amount of the toxic waste and shortening their half lifetime to only a few hundred years. Finally, it solves the Uranium reserve problem because it can use Thorium as its fuel. The Thorium reserve is much larger than that of Uranium. Although the idea of the accelerator-driven nuclear reactor was proposed long time ago, it has not been utilized yet first by technical difficulty and economical reasons. The accelerator-based system needs 1 GeV, 10 MW power proton accelerator. A conventional linear accelerator would need several hundred m length, which is highly costly particularly in Korea because of the high land cost. However, recent technologies make it possible to realize that scale accelerator by a reasonable size. That is the fixed-field alternating gradient (FFAG) accelerator that is described in this article

  2. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  3. Neutron analysis of the fuel of high temperature nuclear reactors; Analisis neutronico del combustible de reactores nucleares de alta temperatura

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    In this work a neutron analysis of the fuel of some high temperature nuclear reactors is presented, studying its main features, besides some alternatives of compound fuel by uranium and plutonium, and of coolant: sodium and helium. For this study was necessary the use of a code able to carry out a reliable calculation of the main parameters of the fuel. The use of the Monte Carlo method was convenient to simulate the neutrons transport in the reactor core, which is the base of the Serpent code, with which the calculations will be made for the analysis. (Author)

  4. A probabilistic safety analysis of incidents in nuclear research reactors.

    Science.gov (United States)

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  5. Software reliability and safety in nuclear reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  6. Passive heat-transfer means for nuclear reactors. [LMFBR

    Science.gov (United States)

    Burelbach, J.P.

    1982-06-10

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  7. The Advanced BWR Nuclear Plant: Safe, economic nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Redding, J.R. [GE Nuclear Energy, San Jose, CA (United States)

    1994-12-31

    The safety and economics of Advanced BWR Nuclear Power Plants are outlined. The topics discussed include: ABWR Programs: status in US and Japan; ABWR competitiveness: safety and economics; SBWR status; combining ABWR and SBWR: the passive ABWR; and Korean/GE partnership.

  8. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1973-01-01

    Advances in Nuclear Science and Technology, Volume 7 provides information pertinent to the fundamental aspects of nuclear science and technology. This book discusses the safe and beneficial development of land-based nuclear power plants.Organized into five chapters, this volume begins with an overview of irradiation-induced void swelling in austenitic stainless steels. This text then examines the importance of various transport processes for fission product redistribution, which depends on the diffusion data, the vaporization properties, and the solubility in the fuel matrix. Other chapters co

  9. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  10. ZEEP: Canada's first nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Green, R.E.; Okazaki, A. [retired, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2015-09-15

    In 1905 Albert Einstein published his historic paper on special relativity, which contained the equation E=mc 2. The significance of this mass-energy relationship became evident with the discovery of nuclear fission in 1939, when it was realized that large amounts of energy would be released in a fission chain reaction. Canadian scientists were involved in this field from the beginning and their efforts resulted in the startup in September 1945 of the ZEEP reactor at Chalk River, the first reactor to go critical outside the USA. In this paper we recall some of the events that led to the construction of ZEEP, and describe the role it played in the development of the Canadian nuclear energy program. (author)

  11. Systems and methods for dismantling a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

    2014-10-28

    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  12. Fission-Produced (99)Mo Without a Nuclear Reactor.

    Science.gov (United States)

    Youker, Amanda J; Chemerisov, Sergey D; Tkac, Peter; Kalensky, Michael; Heltemes, Thad A; Rotsch, David A; Vandegrift, George F; Krebs, John F; Makarashvili, Vakho; Stepinski, Dominique C

    2017-03-01

    (99)Mo, the parent of the widely used medical isotope (99m)Tc, is currently produced by irradiation of enriched uranium in nuclear reactors. The supply of this isotope is encumbered by the aging of these reactors and concerns about international transportation and nuclear proliferation. Methods: We report results for the production of (99)Mo from the accelerator-driven subcritical fission of an aqueous solution containing low enriched uranium. The predominately fast neutrons generated by impinging high-energy electrons onto a tantalum convertor are moderated to thermal energies to increase fission processes. The separation, recovery, and purification of (99)Mo were demonstrated using a recycled uranyl sulfate solution. Conclusion: The (99)Mo yield and purity were found to be unaffected by reuse of the previously irradiated and processed uranyl sulfate solution. Results from a 51.8-GBq (99)Mo production run are presented.

  13. Inception and evolution of Oklo natural nuclear reactors

    Science.gov (United States)

    Bentridi, Salah-Eddine; Gall, Benoît; Gauthier-Lafaye, François; Seghour, Abdeslam; Medjadi, Djamel-Eddine

    2011-11-01

    The occurrence of more than 15 natural nuclear Reactor Zones (RZ) in a geological environment remains a mystery even 40 years after their discovery. The present work gives for the first time an explanation of the chemical and physical processes that caused the start-up of the fission reactions with two opposite processes, uranium enrichments and progressive impoverishment in 235U. Based on Monte-Carlo neutronics simulations, a solution space was defined taking into account realistic combinations of relevant parameters acting on geological conditions and neutron transport physics. This study explains criticality occurrence, operation, expansion and end of life conditions of Oklo natural nuclear reactors, from the smallest to the biggest ones.

  14. Temperature measuring analysis of the nuclear reactor fuel assembly

    Science.gov (United States)

    F., Urban; Ľ., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.

    2014-08-01

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  15. Bibliography of Connecticut Advanced Nuclear Engineering Laboratory reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-01

    This report, published in two volumes, is a bibliography of the reports published at the Connecticut Advanced Nuclear Engineering Laboratory (CANEL). The reports cover the period 1952 through 1965 and include the Aircraft Nuclear Propulsion program, the Advanced Liquid Metal Cooled Reactor program, the Advanced Reactor Materials program and the SNAP-50 program. The bibliography contains the report number, title, author, date published, and classification. In some cases where the writing of a report was a group effort, and in some reports containing compilations of certain types of data, the author column is not applicable. This is indicated by a {open_quotes}n.a.{close_quotes} in the author column. The following types of reports are included: PWAC`s, TIM`s, CNLM`s, FXM`s and miscellaneous reports. PWAC and TIM reports conform to the requirements of AEC Manual Chapter 3202-041 and 3202-042, respectively. Most of the technical information of interest generated by this project is documented in these reports. CNLM and FXM reports were written primarily for internal distribution. However, these reports contain enough information of technical interest to warrant their inclusion. All CNLM`s and those FXM`s considered to be of interest are included in this bibliography. The MPR`s (Monthly Progress Reports) are the most important of the miscellaneous categories of reports. The other miscellaneous categories relate primarily to equipment and reactor specifications. The Division of Technical Information Extension (DTIE) at Oak Ridge, Tennessee has been designated as the primary recipient of the reports in the CANEL library. When more than one copy of a report was available, the additional copies were delivered to the Lawrence Radiation Laboratory, Livermore, California.

  16. Bibliography of Connecticut Advanced Nuclear Engineering Laboratory reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-01

    This report, published in two, volumes, is a bibliography of the reports published at the Connecticut Advanced Nuclear Engineering Laboratory (CANEL). The reports cover the period 1952 through 1965 and include the Aircraft Nuclear Propulsion program, the Advanced Liquid Metal Cooled Reactor program, the Advanced Reactor Materials program and the SNAP-50 program. The bibliography contains the report number, title, author, date published, and classification. In some cases where the writing of a report was a group effort, and in some reports containing compilations of certain types of data, the author column is not applicable. This is indicated by a {open_quotes}n.a.{close_quotes} in the author column. The following types of reports are included: PWAC`s, TIM`s, CNLM`s. FXM`s and miscellaneous reports. PWAC and TIM reports conform to the requirements of AEC Manual Chapter 3202-041 and 3202-042, respectively. Most of the technical information of interest generated by this project is documented in these reports, CNLM and FXM reports were written primarily for internal distribution. However, these reports contain enough information of technical interest to warrant their inclusion. All CNLM`s and those FXM`s considered to be of interest are included in this bibliography. The MPR`s (Monthly Progress Reports) are the most important of the miscellaneous categories of reports. The other miscellaneous categories relate primarily to equipment and reactor specifications. The Division of Technical Information Extension (DTIE) at Oak Ridge, Tennessee has been designated as the primary recipient of the reports in the CANEL library. When more than one copy of a report was available, the additional copies were delivered to the Lawrence Radiation Laboratory, Livermore, California.

  17. Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor

    CERN Document Server

    Int. At. Energy Agency, Wien

    2002-01-01

    This document describes the specific dispositions relative to the nuclear reactor domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor and of the retraining, the Passerelle formation, are presented. (A.L.B.)

  18. 10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

    Science.gov (United States)

    2010-01-01

    ... designed for inserting or removing fuel in an operating nuclear reactor. (3) Complete reactor control rod... contain fuel elements and the primary coolant in a nuclear reactor at an operating pressure in excess of... diffuser plates especially designed or prepared for use in a nuclear reactor. (8) Reactor control......

  19. Towards an efficient multiphysics model for nuclear reactor dynamics

    Directory of Open Access Journals (Sweden)

    Obaidurrahman K.

    2015-01-01

    Full Text Available Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient.

  20. Neutron physics for nuclear reactors unpublished writings by Enrico Fermi

    CERN Document Server

    Fermi, Enrico; Pisanti, O

    2010-01-01

    This unique volume gives an accurate and very detailed description of the functioning and operation of basic nuclear reactors, as emerging from yet unpublished papers by Nobel Laureate Enrico Fermi. In the first part, the entire course of lectures on Neutron Physics delivered by Fermi at Los Alamos is reported, according to the version made by Anthony P French. Here, the fundamental physical phenomena are described very clearly and comprehensively, giving the appropriate physics grounds for the functioning of nuclear piles. In the second part, all the patents issued by Fermi (and coworkers) on

  1. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors; Desarrollo de un program de computo de calculo rapido para el prediseno de celdas de combustible nuclear avanzado 10 x 10 para reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Perusquia, R.; Montes, J.L.; Ortiz, J.J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mrpc@nuclear.inin.mx

    2005-07-01

    In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

  2. Expert system for online surveillance of nuclear reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-12-31

    This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  3. Using ORIGEN2 to Predict Nuclear Reactor Fuel Compositions.

    Science.gov (United States)

    1988-03-01

    Although the principal use of ORIGEN2 is to calculate the isotopic composition of nuclear materials, the following parameters may also be computed with...V’v inal vector by a second vector before storing in the destination vector. BUP: Burnup calculation . Identifies the beginning and end of a series of...has no effect on the accuracy of the calculations . Pressurized Water Reactor (PWR), 33 GWd/MTIHM The ORIGEN2 PWR models are based on a Westinghouse

  4. Determination of 36Cl in nuclear waste from reactor decommissioning

    DEFF Research Database (Denmark)

    Hou, Xiaolin; Frøsig, Lars; Nielsen, Sven Poul

    2007-01-01

    An analytical method for the determination of Cl-36 in nuclear waste such as graphite, heavy concrete, steel, aluminum, and lead was developed. Several methods were investigated for decomposing the samples. AgCl precipitation was used to separate Cl-36 from the matrix elements, followed by ion...... of this analytical method for Cl-36 is 14 mBq. The method has been used to determine Cl-36 in heavy concrete, aluminum, and graphite from the Danish DR-2 research reactor....

  5. Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Higgins, J.; Nasta, K.

    1997-01-01

    The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs.

  6. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  7. Advanced research workshop: nuclear materials safety

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J; Moshkov, M M

    1999-01-28

    The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on the storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of

  8. Neutron dose estimation in a zero power nuclear reactor

    Science.gov (United States)

    Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

    2016-10-01

    This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

  9. Directions for advanced use of nuclear power in century XXI

    Energy Technology Data Exchange (ETDEWEB)

    Walter, C E

    1999-05-01

    Nuclear power can provide a significant contribution to electricity generation and meet other needs of the world and the US during the next century provided that certain directions are taken to achieve its public acceptance. These directions include formulation of projections of population, energy consumption, and energy resources over a responsible period of time. These projections will allow assessment of cumulative effects on the environment and on fixed resources. Use of fossil energy resources in a century of growing demand for energy must be considered in the context of long-term environmental damage and resource depletion. Although some question the validity of these consequences, they can be mitigated by use of advanced fast reactor technology. It must be demonstrated that nuclear power technology is safe, resistant to material diversion for weapon use, and economical. An unbiased examination of all the issues related to energy use, especially of electricity, is an essential direction to take.

  10. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    Science.gov (United States)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  11. A coupled nuclear reactor thermal energy storage system for enhanced load following operation

    Science.gov (United States)

    Alameri, Saeed A.

    Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES

  12. Providing the Basis for Innovative Improvements in Advanced LWR Reactor Passive Safety Systems Design: An Educational R&D Project

    Energy Technology Data Exchange (ETDEWEB)

    Brian G. Williams; Jim C. P. Liou; Hiral Kadakia; Bill Phoenix; Richard R. Schultz

    2007-02-27

    This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems.

  13. Advanced nuclear energy systems and the need of accurate nuclear data: the n_TOF project at CERN

    CERN Document Server

    Colonna, N; Praena, J; Lederer, C; Karadimos, D; Sarmento, R; Domingo-Pardo, C; Plag, R; Massimi, C; Calviani, M; Guerrero, C; Paradela, C; Belloni, F

    2010-01-01

    To satisfy the world's constantly increasing demand for energy, a suitable mix of different energy sources has to be devised. In this scenario, an important role could be played by nuclear energy, provided that major safety, waste and proliferation issues affecting current nuclear reactors are satisfactorily addressed. To this purpose, a large effort has been under way for a few years towards the development of advanced nuclear systems with the aim of closing the fuel cycle. Generation IV reactors, with full or partial waste recycling capability, accelerator driven systems, as well as new fuel cycles are the main options being investigated. The design of advanced systems requires improvements in basic nuclear data, such as cross-sections for neutron-induced reactions on actinides. In this paper, the main concepts of advanced reactor systems are described, together with the related needs of new and accurate nuclear data. The present activity in this field at the neutron facility n\\_TOF at CERN is discussed.

  14. China Advanced Research Reactor Project Progress in 2012

    Institute of Scientific and Technical Information of China (English)

    ZHAO; Tie-jun

    2012-01-01

    <正>In 2012, all the commissioning for the China Advanced Research Reactor (CARR) had been finished and the diffraction pattern had been successfully obtained on the neutron scattering spectrometer. Meanwhile, the cold neutron source project and the acceptance items of CARR project had been carrying out.

  15. The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors

    NARCIS (Netherlands)

    Sjenitzer, B.L.

    2013-01-01

    In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing co

  16. ADVANCED CERAMIC MATERIALS FOR NEXT-GENERATION NUCLEAR APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.

    2010-09-29

    Rising global energy demands coupled with increased environmental concerns point to one solution; they must reduce their dependence on fossil fuels that emit greenhouse gases. As the global community faces the challenge of maintaining sovereign nation security, reducing greenhouse gases, and addressing climate change nuclear power will play a significant and likely growing role. In the US, nuclear energy already provides approximately one-fifth of the electricity used to power factories, offices, homes, and schools with 104 operating nuclear power plants, located at 65 sites in 31 states. Additionally, 19 utilities have applied to the US Nuclear Regulatory Commission (NRC) for construction and operating licenses for 26 new reactors at 17 sites. This planned growth of nuclear power is occurring worldwide and has been termed the 'nuclear renaissance.' As major industrial nations craft their energy future, there are several important factors that must be considered about nuclear energy: (1) it has been proven over the last 40 years to be safe, reliable and affordable (good for Economic Security); (2) its technology and fuel can be domestically produced or obtained from allied nations (good for Energy Security); and (3) it is nearly free of greenhouse gas emissions (good for Environmental Security). Already an important part of worldwide energy security via electricity generation, nuclear energy can also potentially play an important role in industrial processes and supporting the nation's transportation sector. Coal-to-liquid processes, the generation of hydrogen and supporting the growing potential for a greatly increased electric transportation system (i.e. cars and trains) mean that nuclear energy could see dramatic growth in the near future as we seek to meet our growing demand for energy in cleaner, more secure ways. In order to address some of the prominent issues associated with nuclear power generation (i.e., high capital costs, waste management

  17. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  18. Pellet bed reactor concept for nuclear electric propulsion

    Science.gov (United States)

    El-Genk, Mohamed S.; Morley, Nicholas J.; Juhasz, Albert

    1993-01-01

    For Nuclear Electric Propulsion (NEP) applications, gas cooled nuclear reactors with dynamic energy conversion systems offer high specific power and low total mass. This paper describes the Pellet Bed Reactor (PeBR) concept for potential NEP missions to Mars. The helium cooled, 75-80 MWt PeBR, consists of a single annular fuel region filled with a randomly packed bed of spherical fuel pellets, is designed for multiple starts, and offers unique safety and operation features. Each fuel pellet, about 8-10 mm in diameter, is composed of hundreds of TRISO type fuel microspheres embedded in a graphite matrix for a full retention of fission products. To eliminate the likelihood of a single-point failure, the annular core of the PeBR is divided into three 120° sectors. Each sector is self contained and separate and capable of operating and being cooled on its own and in cooperation with either one or two other sectors. Each sector is coupled to a separate, 5 MWe Closed Brayton Cycle (CBC) energy conversion unit and is subcritical for safe handling and launching. In the event of a failure of the cooling system of a core sector, the reactor power level may be reduced, allowing adjacent sectors to convect the heat away using their own cooling system, thus maintaining reactor operation. Also, due to the absence of an internal core structure in the PeBR core, fueling of the reactor can easily be performed either at the launch facility or in orbit, and refueling can be accomplished in orbit as needed to extend the power system lifetime

  19. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  20. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  1. New and recently finalised activities within the NKS Programmes for Nordic cooperation on nuclear reactor safety and emergency preparedness

    DEFF Research Database (Denmark)

    Andgren, Karin; Andersson, Kasper Grann; Magnússon, Sigurður M.

    2015-01-01

    Over the years, NKS has provided funding for hundreds of research activities in fields comprising reactor safety, decommissioning, nuclear and radiological emergency preparedness, and management of radioactive waste. Advanced technologies and methods developed under the NKS framework have been used...... within the Nordic countries as well as internationally. Two programme areas are defined under the NKS platform: The NKS-R programme on nuclear reactor safety and the NKS-B programme on emergency preparedness. Three articles, giving an introduction to NKS and its two programmes, were published...

  2. Investigation of Thermal Hydraulics of a Nuclear Reactor Moderator

    Science.gov (United States)

    Sarchami, Araz

    A three-dimensional numerical modeling of the thermo hydraulics of Canadian Deuterium Uranium (CANDU) nuclear reactor is conducted. The moderator tank is a Pressurized heavy water reactor which uses heavy water as moderator in a cylindrical tank. The main use of the tank is to bring the fast neutrons to the thermal neutron energy levels. The moderator tank compromises of several bundled tubes containing nuclear rods immersed inside the heavy water. It is important to keep the water temperature in the moderator at sub-cooled conditions, to prevent potential failure due to overheating of the tubes. Because of difficulties in measuring flow characteristics and temperature conditions inside a real reactor moderator, tests are conducted using a scaled moderator in moderator test facility (MTF) by Chalk River Laboratories of Atomic Energy of Canada Limited (CRL, AECL). MTF tests are conducted using heating elements to heat tube surfaces. This is different than the real reactor where nuclear radiation is the source of heating which results in a volumetric heating of the heavy water. The data recorded inside the MTF tank have shown levels of fluctuations in the moderator temperatures and requires in depth investigation of causes and effects. The purpose of the current investigation is to determine the causes for, and the nature of the moderator temperature fluctuations using three-dimensional simulation of MTF with both (surface heating and volumetric heating) modes. In addition, three dimensional simulation of full scale actual moderator tank with volumetric heating is conducted to investigate the effects of scaling on the temperature distribution. The numerical simulations are performed on a 24-processor cluster using parallel version of the FLUENT 12. During the transient simulation, 55 points of interest inside the tank are monitored for their temperature and velocity fluctuations with time.

  3. Status and some safety philosophies of the China advanced research reactor CARR

    Energy Technology Data Exchange (ETDEWEB)

    Luzheng Yuan [China Inst. of Atomic Energy, Beijing, BJ (China). Reactor Engineering Research and Design Dept.

    2001-07-01

    The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21{sup st} century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D{sub 2}O reflector not less than 8 x 10{sup 14} n/cm{sup 2}. s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10{sup 15} n/cm{sup 2}. s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

  4. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    Energy Technology Data Exchange (ETDEWEB)

    James E. O' Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  5. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  6. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    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

  7. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  8. Nonlinear Adaptive Dynamic Output-Feedback Power-Level Control of Nuclear Heating Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2013-01-01

    Full Text Available Due to the high safety performance of small nuclear reactors, there is a promising future for small reactors. Nuclear heating reactor (NHR is a small reactor that has many advanced safety features such as the integrated arrangement, natural circulation at any power levels, self-pressurization, hydraulic control rod driving, and passive residual heating removing and can be applied to the fields of district heating, seawater desalination, and electricity production. Since the NHR dynamics has strong nonlinearity and uncertainty, it is meaningful to develop the nonlinear adaptive power-level control technique. From the idea of physically based control design method, a novel nonlinear adaptive power-level control is given for the NHR in this paper. It is theoretically proved that this newly built controller does not only provide globally asymptotic closed-loop stability but is also adaptive to the system uncertainty. Numerical simulation results show the feasibility of this controller and the relationship between the performance and controller parameters.

  9. Seismic base isolation technologies for Korea advanced liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, B.; Lee, J.-H.; Koo, G.-H.; Lee, H.-Y.; Kim, J.-B. [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    2000-06-01

    This paper describes the status and prospects of the seismic base isolation technologies for Korea Advanced Liquid Metal Reactor (KALIMER). The research and development program on the seismic base isolation for KALIMER began in 1993 by KAERI under the national long-term R and D program. The objective of this program is to enhance the seismic safety, to accomplish the economic design, and to standardize the plant design through the establishment of technologies on seismic base isolation for liquid metal reactors. In this paper, tests and analyses performed in the program are presented. (orig.)

  10. Recent Advances on Carbon Molecular Sieve Membranes (CMSMs and Reactors

    Directory of Open Access Journals (Sweden)

    Margot A. Llosa Tanco

    2016-08-01

    Full Text Available Carbon molecular sieve membranes (CMSMs are an important alternative for gas separation because of their ease of manufacture, high selectivity due to molecular sieve separation, and high permeance. The integration of separation by membranes and reaction in only one unit lead to a high degree of process integration/intensification, with associated benefits of increased energy, production efficiencies and reduced reactor or catalyst volume. This review focuses on recent advances in carbon molecular sieve membranes and their applications in membrane reactors.

  11. Commercial US nuclear reactors and waste: the current status

    Energy Technology Data Exchange (ETDEWEB)

    Platt, A.M.; Robinson, J.V.

    1980-09-01

    Between March 1 and June 15, 1980, the declared size of the commercial light waste reactor (LWR) nuclear power industry in the US has decreased another 9 GWe. For the presently declared size: the 165 declared reactors will peak at a capacity of 153 GWe in 2001 and will consume about 870,000 MTU as enrichment feed; the theoretical rate of enrichment requirements will peak at about 19,000,000 SWUs/y in the year 2014; as few as two repositories each with capacity equivalent to 100,000 MTU would hold the waste; and predisposal storage reactor basins and AFRs (away-from-reactor basins) would peak at <85,000 MTU in the year 2020 if the two respositories were commissioned in the years 1997 and 2020. It should be noted that the number of declared LWRs has dropped from 226 on December 31, 1974 to 165 as of this writing. The oil equivalent of the energy loss, assuming a 50% efficiency in use as in cars, is 17,000 million barrels. This is about 10 years of the current rate of US consumption of OPEC oil.

  12. Light weight space power reactors for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H.; Mughabghab, S.; Lazareth, O.; Perkins, K.; Schmidt, E.; Powell, J.R.

    1991-01-01

    A Nuclear Electric Propulsion (NEP) unit capable of propelling a manned vehicle to MARS will be required to have a value of {alpha} (kg/kWe) which is less than five. In order to meet this goal the reactor mass, and thus its contribution to the value of {alpha} will have to be minimized. In this paper a candidate for such a reactor is described. It consists of a gas cooled Particle Bed Reactor (PBR), with specially chosen materials which allow it to operate at an exit temperature of approximately 2000 K. One of the unique features of a PBR is the direct cooling of particulate fuel by the working fluid. This feature allows for high power densities, highest possible gas exit temperatures, for a given fuel temperature and because of the thin particle bed a low pressure drop. The PBR's described in this paper will have a ceramic moderator (Be{sub 2}C), ZrC coated fuel particles and a carbon/carbon hot frit. All the reactors will be designed with sufficient fissile loading to operate at full power for seven years. The burn up possible with particulate fuel is approximately 30%--50%. These rector designs achieve a value of {alpha} less than unity in the power range of interest (5 MWe). 5 refs., 3 figs.

  13. NaK flow control by electromagnetic pump of the SNAP-10A space nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F., E-mail: eduardo@ieav.cta.br, E-mail: fbraz@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAv/DCTA), Sao Jose dos Campos, SP (Brazil)

    2013-07-01

    In the Institute for Advanced Studies (IEAv), were developed, successfully, the first two direct current electromagnetic pumps (DC EM pumps) of Brazil. The first was built with C-type magnet and coils; and the second, with Samarium-Cobalt permanent magnets, for magnetic field generation. Both were tested and performed quite satisfactory. The electromagnetic pump uses the Faraday principle, in which the interaction of the magnetic field and electric current generates the magneto-motive force, which produces the circulation of the fluid. This type of equipment may be used for controlling the liquid metal flow in nuclear space fast reactors. This paper shows the computer programs developed for design and evaluation of DC EM and electromagnetic thermoelectric (EMTE) pumps, the DC EM pump of Samarium-Cobalt magnets data to Mercury loop flow control, the EMTE pump of SNAP space nuclear reactor. It also compares the theoretical results to experimental data of NaK primary loop flow control by electromagnetic thermoelectric pump of the SNAP-10A space nuclear reactor, with satisfactory results, confirming the viability of the electromagnetic pumps evaluation scheme. (author)

  14. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and/or Mars. These reactors require robust automatic control systems using low mass, rapid...

  15. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and Mars. These reactors require robust automatic control systems using low mass, rapid...

  16. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1975-01-01

    Advances in Nuclear Science and Technology, Volume 8 discusses the development of nuclear power in several countries throughout the world. This book discusses the world's largest program of land-based electricity production in the United States.Organized into six chapters, this volume begins with an overview of the phenomenon of quasi-exponential behavior by examining two mathematical models of the neutron field. This text then discusses the finite element method, which is a method for obtaining approximate solutions to integral or differential equations. Other chapters consider the status of

  17. Computational Design of Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Savrasov, Sergey [Univ. of California, Davis, CA (United States); Kotliar, Gabriel [Rutgers Univ., Piscataway, NJ (United States); Haule, Kristjan [Rutgers Univ., Piscataway, NJ (United States)

    2014-06-03

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  18. Qualitative diagnosis for transients analysis on nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lorre, J.P.; Dorlet, E.; Evrard, J.M.

    1995-12-31

    One of the major aims of an intelligent monitoring system, is the supervision task which assist the operator in understanding what occurs on a process. Failures hypotheses must be located and the inferring process must be explained. This paper demonstrate a second generation expert system (SEXTANT) decided to the transients analysis on PWR nuclear reactors. This system detects failures by simulating the process with a numerical model. A diagnosis module uses an even graph built from a causal graph model of the plant to generate hypotheses, and a numerical model to validate these hypotheses. Hypotheses are stored into scenarios which are concurrent possible interpretations of the process evolution. The approach is illustrated by an application for the analysis of the house load operation on a pressurized water reactor. (authors). 9 refs., 10 figs.

  19. SIMODIS - a software package for simulating nuclear reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Guimaraes, Lamartine; Borges, Eduardo M. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados. E-mail: guimarae@ieav.cta.br; Oliveira Junior, Nilton S.; Santos, Glauco S.; Bueno, Mariana F. [Universidade Bras Cubas, Mogi das Cruzes, SP (Brazil)

    2000-07-01

    In this paper it is presented the initial development effort in building a nuclear reactor component simulation package. This package was developed to be used in the MATLAB simulation environment. It uses the graphical capabilities from MATLAB and the advantages of compiled languages, as for instance FORTRAN and C{sup ++}. From the MATLAB it takes the facilities for better displaying the calculated results. From the compiled languages it takes processing speed. So far models from reactor core, UTSG and OTSG have been developed. Also, a series a user-friendly graphical interfaces have been developed for the above models. As a by product a set of water and sodium thermal and physical properties have been developed and may be used directly as a function from MATLAB, or by being called from a model, as part of its calculation process. The whole set was named SIMODIS, which stands for SIstema MODular Integrado de Simulacao. (author)

  20. Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

    Science.gov (United States)

    Carlsen, Robert W.

    Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

  1. Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

    Science.gov (United States)

    Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

    The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

  2. RISMC advanced safety analysis working plan: FY2015 - FY2019. Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H; Smith, Curtis L

    2014-09-01

    In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (“why is this important?”) that will make the case for stakeholder’s use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable “use case” demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

  3. Recent Advances in Ocean Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Kang-Heon Lee

    2015-10-01

    Full Text Available In this paper, recent advances in Ocean Nuclear Power Plants (ONPPs are reviewed, including their general arrangement, design parameters, and safety features. The development of ONPP concepts have continued due to initiatives taking place in France, Russia, South Korea, and the United States. Russia’s first floating nuclear power stations utilizing the PWR technology (KLT-40S and the spar-type offshore floating nuclear power plant designed by a research group in United States are considered herein. The APR1400 and SMART mounted Gravity Based Structure (GBS-type ONPPs proposed by a research group in South Korea are also considered. In addition, a submerged-type ONPP designed by DCNS of France is taken into account. Last, issues and challenges related to ONPPs are discussed and summarized.

  4. Advanced nuclear systems. Review study; Fortgeschrittene Nuklearsysteme. Review Study

    Energy Technology Data Exchange (ETDEWEB)

    Liebert, Wolfgang; Glaser, Alexander; Pistner, Christoph [Interdisziplinaere Arbeitsgruppe Naturwissenschaft, Technik und Sicherheit (IANUS), Darmstadt University of Technology, Hochschulstrasse 10, D-64289 Darmstadt (Germany); Baehr, Roland; Hahn, Lothar [Institute for applied ecology (Oeko-Institut), Elisabethenstrasse 55-57, D-64283 Darmstadt (Germany)

    1999-04-01

    The task of this review study is to from provide an overview of the developments in the field of the various advanced nuclear systems, and to create the basis for more comprehensive studies of technology assessment. In an overview the concepts for advanced nuclear systems pursued worldwide are subdivided into eight subgroups. A coarse examination raster (set pattern) is developed to enable a detailed examination of the selected systems. In addition to a focus on enhanced safety features, further aspects are also taken into consideration, like the lowering of the proliferation risk, the enhancement of the economic competitiveness of the facilities and new usage possibilities (for instance concerning the relaxation of the waste disposal problem or the usage of alternative fuels to uranium). The question about the expected time span for realization and the discussion about the obstacles on the way to a commercially usable reactor also play a substantial role as well as disposal requirements as far as they can be presently recognized. In the central chapter of this study, the documentation of the representatively selected concepts is evaluated as well as existing technology assessment studies and expert opinions. In a few cases where this appears to be necessary, according technical literature, further policy advisory reports, expert statements as well as other relevant sources are taken into account. Contradictions, different assessments and dissents in the literature as well as a few unsettled questions are thus indicated. The potential of advanced nuclear systems with respect to economical and societal as well as environmental objectives cannot exclusively be measured by the corresponding intrinsic or in comparison remarkable technical improvements. The acceptability of novel or improved systems in nuclear technology will have to be judged by their convincing solutions for the crucial questions of safety, nuclear waste and risk of proliferation of nuclear weapons

  5. Nuclear reactor fuel element with vanadium getter on cladding

    Science.gov (United States)

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  6. Evaluation of a hydrogen sensor for nuclear reactor containment monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Hoffheins, B.S.; McKnight, T.E.; Lauf, R.J.; Smith, R.R. [Oak Ridge National Lab., TN (United States); James, R.E. [Electric Power Research Inst., Palo Alto, CA (United States)

    1997-02-01

    Measurement of hydrogen concentration in containment atmospheres in nuclear plants is a key safety capability. Current technologies require extensive sampling systems and subsequent maintenance and calibration costs can be very expensive. A new hydrogen sensor has been developed that is small and potentially inexpensive to install and maintain. Its size and low power requirement make it suitable in distributed systems for pinpointing hydrogen buildup. This paper will address the first phase of a testing program conducted to evaluate this sensor for operation in reactor containments.

  7. Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR)

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, B.G.; Richards, R.E.; Reece, W.J.; Gertman, D.I.

    1992-10-01

    This Reference Guide contains instructions on how to install and use Version 3.5 of the NRC-sponsored Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The NUCLARR data management system is contained in compressed files on the floppy diskettes that accompany this Reference Guide. NUCLARR is comprised of hardware component failure data (HCFD) and human error probability (HEP) data, both of which are available via a user-friendly, menu driven retrieval system. The data may be saved to a file in a format compatible with IRRAS 3.0 and commercially available statistical packages, or used to formulate log-plots and reports of data retrieval and aggregation findings.

  8. Thermo-magnetic systems for space nuclear reactors an introduction

    CERN Document Server

    Maidana, Carlos O

    2014-01-01

    Introduces the reader to engineering magnetohydrodynamics applications and presents a comprehensive guide of how to approach different problems found in this multidisciplinary field. An introduction to engineering magnetohydrodynamics, this brief focuses heavily on the design of thermo-magnetic systems for liquid metals, with emphasis on the design of electromagnetic annular linear induction pumps for space nuclear reactors. Alloy systems that are liquid at room temperature have a high degree of thermal conductivity far superior to ordinary non-metallic liquids. This results in their use for

  9. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  10. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  11. Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    J. Wesley Hines; Belle R. Upadhyaya; J. Michael Doster; Robert M. Edwards; Kenneth D. Lewis; Paul Turinsky; Jamie Coble

    2011-05-31

    Development and deployment of small-scale nuclear power reactors and their maintenance, monitoring, and control are part of the mission under the Small Modular Reactor (SMR) program. The objectives of this NERI-consortium research project are to investigate, develop, and validate advanced methods for sensing, controlling, monitoring, diagnosis, and prognosis of these reactors, and to demonstrate the methods with application to one of the proposed integral pressurized water reactors (IPWR). For this project, the IPWR design by Westinghouse, the International Reactor Secure and Innovative (IRIS), has been used to demonstrate the techniques developed under this project. The research focuses on three topical areas with the following objectives. Objective 1 - Develop and apply simulation capabilities and sensitivity/uncertainty analysis methods to address sensor deployment analysis and small grid stability issues. Objective 2 - Develop and test an autonomous and fault-tolerant control architecture and apply to the IRIS system and an experimental flow control loop, with extensions to multiple reactor modules, nuclear desalination, and optimal sensor placement strategy. Objective 3 - Develop and test an integrated monitoring, diagnosis, and prognosis system for SMRs using the IRIS as a test platform, and integrate process and equipment monitoring (PEM) and process and equipment prognostics (PEP) toolboxes. The research tasks are focused on meeting the unique needs of reactors that may be deployed to remote locations or to developing countries with limited support infrastructure. These applications will require smaller, robust reactor designs with advanced technologies for sensors, instrumentation, and control. An excellent overview of SMRs is described in an article by Ingersoll (2009). The article refers to these as deliberately small reactors. Most of these have modular characteristics, with multiple units deployed at the same plant site. Additionally, the topics focus

  12. Recent Advances in Pd-Based Membranes for Membrane Reactors.

    Science.gov (United States)

    Arratibel Plazaola, Alba; Pacheco Tanaka, David Alfredo; Van Sint Annaland, Martin; Gallucci, Fausto

    2017-01-01

    Palladium-based membranes for hydrogen separation have been studied by several research groups during the last 40 years. Much effort has been dedicated to improving the hydrogen flux of these membranes employing different alloys, supports, deposition/production techniques, etc. High flux and cheap membranes, yet stable at different operating conditions are required for their exploitation at industrial scale. The integration of membranes in multifunctional reactors (membrane reactors) poses additional demands on the membranes as interactions at different levels between the catalyst and the membrane surface can occur. Particularly, when employing the membranes in fluidized bed reactors, the selective layer should be resistant to or protected against erosion. In this review we will also describe a novel kind of membranes, the pore-filled type membranes prepared by Pacheco Tanaka and coworkers that represent a possible solution to integrate thin selective membranes into membrane reactors while protecting the selective layer. This work is focused on recent advances on metallic supports, materials used as an intermetallic diffusion layer when metallic supports are used and the most recent advances on Pd-based composite membranes. Particular attention is paid to improvements on sulfur resistance of Pd based membranes, resistance to hydrogen embrittlement and stability at high temperature.

  13. Recent Advances in Pd-Based Membranes for Membrane Reactors

    Directory of Open Access Journals (Sweden)

    Alba Arratibel Plazaola

    2017-01-01

    Full Text Available Palladium-based membranes for hydrogen separation have been studied by several research groups during the last 40 years. Much effort has been dedicated to improving the hydrogen flux of these membranes employing different alloys, supports, deposition/production techniques, etc. High flux and cheap membranes, yet stable at different operating conditions are required for their exploitation at industrial scale. The integration of membranes in multifunctional reactors (membrane reactors poses additional demands on the membranes as interactions at different levels between the catalyst and the membrane surface can occur. Particularly, when employing the membranes in fluidized bed reactors, the selective layer should be resistant to or protected against erosion. In this review we will also describe a novel kind of membranes, the pore-filled type membranes prepared by Pacheco Tanaka and coworkers that represent a possible solution to integrate thin selective membranes into membrane reactors while protecting the selective layer. This work is focused on recent advances on metallic supports, materials used as an intermetallic diffusion layer when metallic supports are used and the most recent advances on Pd-based composite membranes. Particular attention is paid to improvements on sulfur resistance of Pd based membranes, resistance to hydrogen embrittlement and stability at high temperature.

  14. Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

    2013-07-01

    In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

  15. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  16. Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-04-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

  17. A cermet fuel reactor for nuclear thermal propulsion

    Science.gov (United States)

    Kruger, Gordon

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

  18. Computer simulation of two-phase flow in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.

    1992-09-01

    Two-phase flow models dominate the economic resource requirements for development and use of computer codes for analyzing thermohydraulic transients in nuclear power plants. Six principles are presented on mathematical modeling and selection of numerical methods, along with suggestions on programming and machine selection, all aimed at reducing the cost of analysis. Computer simulation is contrasted with traditional computer calculation. The advantages of run-time interactive access operation in a simulation environment are demonstrated. It is explained that the drift-flux model is better suited for two-phase flow analysis in nuclear reactors than the two-fluid model, because of the latter`s closure problem. The advantage of analytical over numerical integration is demonstrated. Modeling and programming techniques are presented which minimize the number of needed arithmetical and logical operations and thereby increase the simulation speed, while decreasing the cost.

  19. Computer simulation of two-phase flow in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.

    1992-01-01

    Two-phase flow models dominate the economic resource requirements for development and use of computer codes for analyzing thermohydraulic transients in nuclear power plants. Six principles are presented on mathematical modeling and selection of numerical methods, along with suggestions on programming and machine selection, all aimed at reducing the cost of analysis. Computer simulation is contrasted with traditional computer calculation. The advantages of run-time interactive access operation in a simulation environment are demonstrated. It is explained that the drift-flux model is better suited for two-phase flow analysis in nuclear reactors than the two-fluid model, because of the latter's closure problem. The advantage of analytical over numerical integration is demonstrated. Modeling and programming techniques are presented which minimize the number of needed arithmetical and logical operations and thereby increase the simulation speed, while decreasing the cost.

  20. Recent Advances on Carbon Molecular Sieve Membranes (CMSMs) and Reactors

    OpenAIRE

    Margot A. Llosa Tanco; Pacheco Tanaka, David A.

    2016-01-01

    Carbon molecular sieve membranes (CMSMs) are an important alternative for gas separation because of their ease of manufacture, high selectivity due to molecular sieve separation, and high permeance. The integration of separation by membranes and reaction in only one unit lead to a high degree of process integration/intensification, with associated benefits of increased energy, production efficiencies and reduced reactor or catalyst volume. This review focuses on recent advances in carbon mole...

  1. Nordic Nuclear Materials Forum for Generation IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anghel, C. (Studsvik Nuclear AB, Nykoeping (Sweden)); Penttilae, S. (Technical Research Centre of Finland, VTT (Finland))

    2010-03-15

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  2. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  3. High-intensity power-resolved radiation imaging of an operational nuclear reactor

    Science.gov (United States)

    Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

    2015-01-01

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

  4. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    T. Höhne

    2010-01-01

    Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

  5. RTC-control of power transients in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, Wajdi Mohamed [Alfateh University, PO Box 13040, Tripoli (Libyan Arab Jamahiriya)

    2006-07-01

    In this paper, the new Reactivity Trace Curve (RTC) method (Ratemi 1993,1994), which is based on the dynamic period studies (Bernard et al.,1984), has been studied for maneuvering of the nuclear reactor power without power shooting. The reactor is modeled with one group of delayed neutrons with temperature feedback effect, as well as, Xenon feedback effect. A precursors concentration model is used to provide for the effective dynamic decay constant (in one group case, it is a static one). The RTC-identifier which is given by a differential equation is then solved at each sampling time (for one group, it has an analytical solution). Its solution is what is called the Reactivity Trace Curve which keeps the power steady at the desired power. An inverse kinetic model which uses the on-line power data for reactivity calculation is used to provide initial condition (initial reactivity) for the RTC- power controller. Also feedback model are needed to evaluate both the temperature and Xenon reactivities which when subtracted from the RTC-value, one then can determine the reactivity required to keep the reactor power steady without power shooting. (authors)

  6. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Fractional neutron point kinetics equations for nuclear reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Paredes, Gilberto, E-mail: gepe@xanum.uam.mx [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, Mexico, D.F. 09340 (Mexico); Polo-Labarrios, Marco-A. [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, Mexico, D.F. 09340 (Mexico); Espinosa-Martinez, Erick-G. [Retorno Quebec 6, Col. Burgos de Cuernavaca 62580, Temixco, Mor. (Mexico); Valle-Gallegos, Edmundo del [Escuela Superior de Fisica y Matematicas, Instituto Politecnico Nacional, Av. Instituto Politecnico Nacional s/n, Col. San Pedro Zacatenco, Mexico, D.F. 07738 (Mexico)

    2011-02-15

    The fractional point-neutron kinetics model for the dynamic behavior in a nuclear reactor is derived and analyzed in this paper. The fractional model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional order, acting as exponent of the relaxation time, to obtain the best representation of a nuclear reactor dynamics. The physical interpretation of the fractional order is related with non-Fickian effects from the neutron diffusion equation point of view. The numerical approximation to the solution of the fractional neutron point kinetics model, which can be represented as a multi-term high-order linear fractional differential equation, is calculated by reducing the problem to a system of ordinary and fractional differential equations. The numerical stability of the fractional scheme is investigated in this work. Results for neutron dynamic behavior for both positive and negative reactivity and for different values of fractional order are shown and compared with the classic neutron point kinetic equations. Additionally, a related review with the neutron point kinetics equations is presented, which encompasses papers written in English about this research topic (as well as some books and technical reports) published since 1940 up to 2010.

  9. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  10. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  11. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  12. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  13. Advances in Process Intensification through Multifunctional Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    O' Hern, Timothy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center; Evans, Lindsay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Sciences and Engineering Center; Miller, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Materials Sciences and Engineering Center; Cooper, Marcia [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Energetic Components Realization Center; Torczynski, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pena, Donovan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gill, Walt [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Engineering Sciences Center

    2011-02-01

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in

  14. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  15. Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Utgikar, Vivek [Univ. of Idaho, Moscow, ID (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Christensen, Richard [The Ohio State Univ., Columbus, OH (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-29

    The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate the models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.

  16. Sites for locations of nuclear reactors; Sitios para emplazamientos de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Balcazar, M.; Huerta, M.; Lopez, A., E-mail: miguel.balcazar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    A restriction on sites of nuclear energy is the history of seismic activity, in its magnitude (Richter) and intensity (Mercalli). This article delimits the areas of greatest magnitude and national seismic intensity, with restrictions of ground acceleration; the supplement areas with a low magnitude of seismic activity are shown. Potential sites for the location of these sites are introduced into a geographic information system. The set of geo-referenced data contains the location of the active volcanic manifestations; the historical record of earthquake epicenters, magnitudes and intensities; major geological faults; surface hydrology and water bodies; location of population density; protected areas; contour lines; the rock type or geology. The geographic information system allows entering normative criteria and environmental restrictions that correlate with geo-referenced data described above, forms both probable and exclusion areas for the installation of nuclear sites. (Author)

  17. A Spouted Bed Reactor Monitoring System for Particulate Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    D. S. Wendt; R. L. Bewley; W. E. Windes

    2007-06-01

    Conversion and coating of particle nuclear fuel is performed in spouted (fluidized) bed reactors. The reactor must be capable of operating at temperatures up to 2000°C in inert, flammable, and coating gas environments. The spouted bed reactor geometry is defined by a graphite retort with a 2.5 inch inside diameter, conical section with a 60° included angle, and a 4 mm gas inlet orifice diameter through which particles are removed from the reactor at the completion of each run. The particles may range from 200 µm to 2 mm in diameter. Maintaining optimal gas flow rates slightly above the minimum spouting velocity throughout the duration of each run is complicated by the variation of particle size and density as conversion and/or coating reactions proceed in addition to gas composition and temperature variations. In order to achieve uniform particle coating, prevent agglomeration of the particle bed, and monitor the reaction progress, a spouted bed monitoring system was developed. The monitoring system includes a high-sensitivity, low-response time differential pressure transducer paired with a signal processing, data acquisition, and process control unit which allows for real-time monitoring and control of the spouted bed reactor. The pressure transducer is mounted upstream of the spouted bed reactor gas inlet. The gas flow into the reactor induces motion of the particles in the bed and prevents the particles from draining from the reactor due to gravitational forces. Pressure fluctuations in the gas inlet stream are generated as the particles in the bed interact with the entering gas stream. The pressure fluctuations are produced by bulk movement of the bed, generation and movement of gas bubbles through the bed, and the individual motion of particles and particle subsets in the bed. The pressure fluctuations propagate upstream to the pressure transducer where they can be monitored. Pressure fluctuation, mean differential pressure, gas flow rate, reactor

  18. Advancing Underground Nuclear Astrophysics with CASPAR

    Science.gov (United States)

    Robertson, Daniel; Couder, Manoel; Greife, Uwe; Strieder, Frank; Wells, Doug; Wiescher, Michael

    2015-04-01

    The advancement of experimental nuclear astrophysics techniques and the requirement of astrophysical network models for further nuclear data over greater energy ranges, has led to the requirement for the better understanding of nuclear reactions in stellar burning regimes. For those reactions of importance to stellar burning processes and elemental production through stellar nucleosynthesis, the energy range of astrophysical interest is always problematic to probe. As reaction measurements approach the burning window of interest, the rapid drop off in cross-section hampers laboratory investigation. The natural background suppression of underground accelerator facilities enables the extension of current experimental data to lower energies. An example of such reactions of interest are those thought to be sources of neutrons for the s-process, the major production mechanism for elements above the iron peak. The reactions 13 C(α,n)16 O and 22 Ne(α,n)25 Mg are the proposed initial focus of the new nuclear astrophysics accelerator laboratory (CASPAR) currently under construction at the Sanford Underground Research Facility, Lead, SD. With thanks to funding provided by South Dakota Science and Technology Authority and the NSF under Grant Number PHY-1419765.

  19. Development and Assessment of Advanced Reactor Core Protection System

    Science.gov (United States)

    in, Wang-Kee; Park, Young-Ho; Baeg, Seung-Yeob

    An advanced core protection system for a pressurized water reactor, Reactor Core Protection System(RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%.

  20. Deterministic and risk-informed approaches for safety analysis of advanced reactors: Part II, Risk-informed approaches

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Inn Seock, E-mail: innseockkim@gmail.co [ISSA Technology, 21318 Seneca Crossing Drive, Germantown, MD 20876 (United States); Ahn, Sang Kyu; Oh, Kyu Myung [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)

    2010-05-15

    Technical insights and findings from a critical review of deterministic approaches typically applied to ensure design safety of nuclear power plants were presented in the companion paper of Part I included in this issue. In this paper we discuss the risk-informed approaches that have been proposed to make a safety case for advanced reactors including Generation-IV reactors such as Modular High-Temperature Gas-cooled Reactor (MHTGR), Pebble Bed Modular Reactor (PBMR), or Sodium-cooled Fast Reactor (SFR). Also considered herein are a risk-informed safety analysis approach suggested by Westinghouse as a means to improve the conventional accident analysis, together with the Technology Neutral Framework recently developed by the US Nuclear Regulatory Commission as a high-level regulatory infrastructure for safety evaluation of any type of reactor design. The insights from a comparative review of various deterministic and risk-informed approaches could be usefully used in developing a new licensing architecture for enhanced safety of evolutionary or advanced plants.

  1. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

    Energy Technology Data Exchange (ETDEWEB)

    Whitehead, Donnie Wayne; Varnado, G. Bruce

    2008-09-01

    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilistic risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.

  2. Enhanced in-pile instrumentation at the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

  3. Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    William Richins; Stephen Novascone; Cheryl O' Brien

    2009-08-01

    Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors William Richins1, Stephen Novascone1, and Cheryl O’Brien1 1Idaho National Laboratory, US Dept. of Energy, Idaho Falls, Idaho, USA, e-mail: William.Richins@inl.gov The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selected to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; 1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and 2) calibrating simulation software and methods that address topic 1 The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.

  4. Protactinium-231 as a new fissionable material for nuclear reactors that can produce nuclear fuel with stable neutron-multiplying properties

    Energy Technology Data Exchange (ETDEWEB)

    Shmelev, Anatoly N.; Kulikov, Gennady G.; Kulikov, Evgeny G.; Apse, Vladimir A. [National Research Nuclear Univ. MEPHI, Moscow (Russian Federation). Moscow Engineering Physics Inst.

    2016-03-15

    Main purpose of the study is justifying doping of protactinium-231 into fuel compositions of advanced nuclear reactors with the ultimate aim to improve their operation safety and economic efficiency. Protactinium-231 could be generated in thorium blankets of hybrid thermonuclear facilities. The following results were obtained: 1. Protactinium-231 has some favorable features for its doping into nuclear fuel; 2. Protactinium containing fuel compositions can be characterized by the higher values of fuel burn-up, the longer values of fuel lifetime and the better proliferation resistance; 3. as protactinium-231 is the stronger neutron absorber than uranium-238, remarkably lower amounts of protactinium-231 may be doped into fuel compositions. The free space could be occupied by materials which are able to improve heat conductivity and refractoriness of fuel. As a consequence, operation safety of nuclear reactors could be upgraded.

  5. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  6. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  7. Fiscal year 1999 multi-year work plan, advanced reactors transition program

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D.A.

    1998-09-17

    The Advanced Reactors Transition (ART) has two missions. One, funded by DOE-EM is to transition assigned, surplus facilities to a safe and compliant, low-cost stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D and D. Facilities to be transitioned include the 309 Building/Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy (NE) Legacy Facilities. The second mission, funded by DOE-NE, is to maintain the Fast Flux Test Facility (FFTF) and affiliated 400 Area buildings in a safe and compliant standby condition. The condition of the plant hardware, software and personnel is to be preserved in a manner not to preclude a plant restart.

  8. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  9. Heterogeneous Nuclear Reactor Models for Optimal Xenon Control.

    Science.gov (United States)

    Gondal, Ishtiaq Ahmad

    Nuclear reactors are generally modeled as homogeneous mixtures of fuel, control, and other materials while in reality they are heterogeneous-homogeneous configurations comprised of fuel and control rods along with other materials. Similarly, for space-time studies of a nuclear reactor, homogeneous, usually one-group diffusion theory, models are used, and the system equations are solved by either nodal or modal expansion approximations. Study of xenon-induced problems has also been carried out using similar models and with the help of dynamic programming or classical calculus of variations or the minimum principle. In this study a thermal nuclear reactor is modeled as a two-dimensional lattice of fuel and control rods placed in an infinite-moderator in plane geometry. The two-group diffusion theory approximation is used for neutron transport. Space -time neutron balance equations are written for two groups and reduced to one space-time algebraic equation by using the two-dimensional Fourier transform. This equation is written at all fuel and control rod locations. Iodine -xenon and promethium-samarium dynamic equations are also written at fuel rod locations only. These equations are then linearized about an equilibrium point which is determined from the steady-state form of the original nonlinear system equations. After studying poisonless criticality, with and without control, and the stability of the open-loop system and after checking its controllability, a performance criterion is defined for the xenon-induced spatial flux oscillation problem in the form of a functional to be minimized. Linear -quadratic optimal control theory is then applied to solve the problem. To perform a variety of different additional useful studies, this formulation has potential for various extensions and variations; for example, different geometry of the problem, with possible extension to three dimensions, heterogeneous -homogeneous formulation to include, for example, homogeneously

  10. Joining of Oxide Dispersion Strengthened Steels for Advanced Reactors

    Science.gov (United States)

    Baker, B. W.; Brewer, L. N.

    2014-12-01

    The design, manufacture, and experimental analysis of structural materials capable of operation in the high temperatures, corrosive environments, and radiation damage spectra of future reactor designs remain one of the key pacing items for advanced reactor designs. The most promising candidate structural materials are vanadium-based refractory alloys, silicon carbide composites and oxide dispersion strengthened steels. Of these, oxide dispersion strengthened steels are a likely near-term candidate to meet required demands. This paper reviews different variants of oxide dispersion strengthened steels and discusses their capability with regard to high-temperature strength, corrosion resistance, and radiation damage resistance. Additionally, joining of oxide dispersion strengthened steels, which has been cited as a limiting factor preventing their use, is addressed and reviewed. Specifically, friction stir welding of these steels is reviewed as a promising joining method for oxide dispersion strengthened steels.

  11. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  12. Nondestructive Measurements for Diagnostics of Advanced Reactor Passive Components

    Energy Technology Data Exchange (ETDEWEB)

    Prowant, Matthew S.; Dib, Gerges; Roy, Surajit; Luzi, Lorenzo; Ramuhalli, Pradeep

    2016-09-20

    Information on advanced reactor (AdvRx) component condition and failure probability is necessary to maintaining adequate safety margins and avoiding unplanned shutdowns, both of which have regulatory and economic consequences. Prognostic health management (PHM) technologies provide one approach to addressing these needs by providing the technical means for lifetime management of significant passive components and reactor internals. However, such systems require measurement data that are sensitive to degradation of the component. This paper describes results to date of ongoing research on nondestructive measurements of component condition for degradation mechanisms of relevance to AdvRx concepts. The focus of this paper is on in-situ ultrasonic measurements during high-temperature creep degradation. The data were analyzed to assess the sensitivity of the measurements to creep degradation, with the specific objective of assessing the suitability of the resulting correlations for remaining life prediction. The details of the measurements, results of data analysis, and ongoing research in this area are discussed.

  13. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  14. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF NUCLEAR EQUIPMENT AND MATERIAL Licenses § 110.26 General license for the export of nuclear...

  15. A Matrix Method of Analyzing the Thermodynamic System of Advance Boiling Water Reactor Nuclear Power Unit%先进型沸水堆核电机组热经济性矩阵分析方法

    Institute of Scientific and Technical Information of China (English)

    冉鹏; 李庚生; 廖丹; 朱伟平

    2010-01-01

    根据先进型沸水堆(advance boiling water reactor,ABWR)核电机组热力系统的结构特点,基于热力系统等效热降分析方法和矩阵方法,确定其主、辅系统的划分原则以及辅助汽水成分划分原则,对先进型沸水堆各种汽水成分进行归并处理,构建表达规则的先进型沸水堆核电机组汽水分布方程填写规则,推导出适合先进型沸水堆核电机组热力系统热经济性分析的通用矩阵方法,并给出该类型核电机组辅助汽水成分对热经济性影响的表达方式.该矩阵全面反映了先进犁沸水堆核电机组热力系统主系统和各种辅助系统对机组热经济性的影响状况,每个子矩阵物理意义明确、规律性强,可使先进型沸水堆核电机组热力系统的整体计算和局部分析变得清晰、简单,适合于计算机程序化,并通过实例对该方法进行了验证.

  16. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    2004-07-29

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  17. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  18. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  19. Measuring of fissile isotopes partial antineutrino spectra in direct experiment at nuclear reactor

    CERN Document Server

    Sinev, V V

    2009-01-01

    The direct measuring method is considered to get nuclear reactor antineutrino spectrum. We suppose to isolate partial spectra of the fissile isotopes by using the method of antineutrino spectrum extraction from the inverse beta decay positron spectrum applied at Rovno experiment. This admits to increase the accuracy of partial antineutrino spectra forming the total nuclear reactor spectrum. It is important for the analysis of the reactor core fuel composition and could be applied for non-proliferation purposes.

  20. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    OpenAIRE

    Sungjoo Lee; Byungun Yoon; Juneseuk Shin

    2016-01-01

    We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indic...

  1. Advancement of safeguards inspection technology for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Sung; Park, W. S.; Cha, H. R.; Ham, Y. S.; Lee, Y. G.; Kim, K. P.; Hong, Y. D

    1999-04-01

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  2. Determination of the fission coefficients in thermal nuclear reactors for antineutrino detection

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Lenilson M. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Cabral, Ronaldo G., E-mail: rgcabral@ime.eb.b [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Anjos, Joao C.C. dos, E-mail: janjos@cbpf.b [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil). Dept. GLN - G

    2011-07-01

    The nuclear reactors in operation periodically need to change their fuel. It is during this process that these reactors are more vulnerable to occurring of several situations of fuel diversion, thus the monitoring of the nuclear installations is indispensable to avoid events of this nature. Considering this fact, the most promissory technique to be used for the nuclear safeguard for the nonproliferation of nuclear weapons, it is based on the detection and spectroscopy of antineutrino from fissions that occur in the nuclear reactors. The detection and spectroscopy of antineutrino, they both depend on the single contribution for the total number of fission of each actinide in the core reactor, these contributions receive the name of fission coefficients. The goal of this research is to show the computational and mathematical modeling used to determinate these coefficients for PWR reactors. (author)

  3. Risks of nuclear energy technology safety concepts of light water reactors

    CERN Document Server

    Kessler, Günter; Schlüter, Franz-Hermann

    2014-01-01

    The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on?reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: ? A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Ch

  4. Determination of 36Cl in nuclear waste from reactor decommissioning.

    Science.gov (United States)

    Hou, Xiaolin; Ostergaard, Lars Frøsig; Nielsen, Sven P

    2007-04-15

    An analytical method for the determination of 36Cl in nuclear waste such as graphite, heavy concrete, steel, aluminum, and lead was developed. Several methods were investigated for decomposing the samples. AgCl precipitation was used to separate 36Cl from the matrix elements, followed by ion-exchange chromatography to remove interfering radionuclides. The purified 36Cl was then measured by liquid scintillation counting. The chemical yield of chlorine, as measured by ICPMS, is above 70% and the decontamination factors for all interfering radionuclides are greater than 10(6). The detection limit of this analytical method for 36Cl is 14 mBq. The method has been used to determine 36Cl in heavy concrete, aluminum, and graphite from the Danish DR-2 research reactor.

  5. Preloading of bolted connections in nuclear reactor component supports

    Energy Technology Data Exchange (ETDEWEB)

    Yahr, G T

    1984-10-01

    A number of failures of threaded fasteners in nuclear reactor component supports have been reported. Many of those failures were attributed to stress corrosion cracking. This report discusses how stress corrosion cracking can be avoided in bolting by controlling the maximum bolt preloads so that the sustained stresses in the bolts are below the level required to cause stress corrosion cracking. This is a basic departure from ordinary bolted joint design where the only limits on preload are on the minimum preload. Emphasis is placed on the importance of detailed analysis to determine the acceptable range of preload and the selection of a method for measuring the preload that is sufficiently accurate to ensure that the preload is actually within the acceptable range. Procedures for determining acceptable preload range are given, and the accuracy of various methods of measuring preload is discussed.

  6. Search for neutrino oscillations at the palo verde nuclear reactors

    Science.gov (United States)

    Boehm; Busenitz; Cook; Gratta; Henrikson; Kornis; Lawrence; Lee; McKinny; Miller; Novikov; Piepke; Ritchie; Tracy; Vogel; Wang; Wolf

    2000-04-24

    We report on the initial results from a measurement of the antineutrino flux and spectrum at a distance of about 800 m from the three reactors of the Palo Verde Nuclear Generating Station using a segmented gadolinium-loaded scintillation detector. We find that the antineutrino flux agrees with that predicted in the absence of oscillations excluding at 90% C.L. nu;(e)-nu;(x) oscillations with Deltam(2)>1.12x10(-3) eV(2) for maximal mixing and sin (2)2straight theta>0.21 for large Deltam(2). Our results support the conclusion that the atmospheric neutrino oscillations observed by Super-Kamiokande do not involve nu(e).

  7. Measurements of the subcriticality using advanced technique of shooting source during operation of NPP reactors

    Science.gov (United States)

    Lebedev, G. V.; Petrov, V. V.; Bobylyov, V. T.; Butov, R. I.; Zhukov, A. M.; Sladkov, A. A.

    2014-12-01

    According to the rules of nuclear safety, the measurements of the subcriticality of reactors should be carried out in the process of performing nuclear hazardous operations. An advanced technique of shooting source of neutrons is proposed to meet this requirement. As such a source, a pulsed neutron source (PNS) is used. In order to realize this technique, it is recommended to enable a PNS with a frequency of 1-20 Hz. The PNS is stopped after achieving a steady-state (on average) number of neutrons in the reactor volume. The change in the number of neutrons in the reactor volume is measured in time with an interval of discreteness of ˜0.1 s. The results of these measurements with the application of a system of point-kinetics equations are used in order to calculate the sought subcriticality. The basic idea of the proposed technique used to measure the subcriticality is elaborated in a series of experiments on the Kvant assembly. The conditions which should be implemented in order to obtain a positive result of measurements are formulated. A block diagram of the basic version of the experimental setup is presented, whose main element is a pulsed neutron generator.

  8. Multiphysics modeling of porous CRUD deposits in nuclear reactors

    Science.gov (United States)

    Short, M. P.; Hussey, D.; Kendrick, B. K.; Besmann, T. M.; Stanek, C. R.; Yip, S.

    2013-11-01

    The formation of porous CRUD deposits on nuclear reactor fuel rods, a longstanding problem in the operation of pressurized water reactors (PWRs), is a significant challenge to science-based multiscale modeling and simulation. While existing, published studies have focused on individual or loosely coupled processes, such as heat transfer, fluid flow, and compound dissolution/precipitation, none have addressed their coupled effects sufficiently to enable a comprehensive, scientific understanding of CRUD. Here we present the formulation and results of a model, MAMBA-BDM, which begins to incorporate mechanistic details in describing CRUD in PWRs. CRUD is treated as a chemical deposition process in an environment of variable concentration, an arbitrary level of heating, and a complex fractal-based flow geometry. We present results on spatial distributions of temperature, pressure, velocity, and concentration that give insight into the interplay between these physical properties and geometrical parameters. We show the role of heat convection which has not been discussed previously. Furthermore, we suggest that the assumption of liquid saturation in the CRUD deserves scrutiny, as a result of our attempt to determine an effective CRUD thermal conductivity.

  9. SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

    Directory of Open Access Journals (Sweden)

    WOLFGANG HARTMANN

    2013-10-01

    Full Text Available This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

  10. High Level Requirements for the Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

    Energy Technology Data Exchange (ETDEWEB)

    Rich Johnson; Hyung Lee; Kimberlyn C. Mousseau

    2011-09-01

    The US Department of Energy, Office of Nuclear Energy (DOE-NE), has been tasked with the important mission of ensuring that nuclear energy remains a compelling and viable energy source in the U.S. The motivations behind this mission include cost-effectively meeting the expected increases in the power needs of the country, reducing carbon emissions and reducing dependence on foreign energy sources. In the near term, to ensure that nuclear power remains a key element of U.S. energy strategy and portfolio, the DOE-NE will be working with the nuclear industry to support safe and efficient operations of existing nuclear power plants. In the long term, to meet the increasing energy needs of the U.S., the DOE-NE will be investing in research and development (R&D) and working in concert with the nuclear industry to build and deploy new, safer and more efficient nuclear power plants. The safe and efficient operations of existing nuclear power plants and designing, licensing and deploying new reactor designs, however, will require focused R&D programs as well as the extensive use and leveraging of advanced modeling and simulation (M&S). M&S will play a key role in ensuring safe and efficient operations of existing and new nuclear reactors. The DOE-NE has been actively developing and promoting the use of advanced M&S in reactor design and analysis through its R&D programs, e.g., the Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Consortium for Advanced Simulation of Light Water Reactors (CASL) programs. Also, nuclear reactor vendors are already using CFD and CSM, for design, analysis, and licensing. However, these M&S tools cannot be used with confidence for nuclear reactor applications unless accompanied and supported by verification and validation (V&V) and uncertainty quantification (UQ) processes and procedures which provide quantitative measures of uncertainty for specific applications. The Nuclear Energy Knowledge base for Advanced Modeling and Simulation

  11. A world class nuclear research reactor complex for South Africa's nuclear future

    Energy Technology Data Exchange (ETDEWEB)

    Keshaw, Jeetesh [South African Young Nuclear Professional Society, PO Box 9396, Centurion, 0157 (South Africa)

    2008-07-01

    South Africa recently made public its rather ambitious goals pertaining to nuclear energy developments in a Draft Policy and Strategy issued for public comment. Not much attention was given to an important tool for nuclear energy research and development, namely a well equipped and maintained research reactor, which on its own does not do justice to its potential, unless it is fitted with all the ancillaries and human resources as most first world countries have. In South Africa's case it is suggested to establish at least one Nuclear Energy Research and Development Centre at such a research reactor, where almost all nuclear energy related research can be carried out on par with some of the best in the world. The purpose of this work is to propose how this could be done, and motivate why it is important that it be done with great urgency, and with full involvement of young professionals, if South Africa wishes to face up to the challenges mentioned in the Draft Strategy and Policy. (authors)

  12. Reference Operational Concepts for Advanced Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques Victor [Idaho National Lab. (INL), Idaho Falls, ID (United States); Farris, Ronald Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This report represents the culmination of a four-year research project that was part of the Instrumentation and Control and Human Machine Interface subprogram of the DOE Advanced Reactor Technologies program.

  13. Radiochemical analysis of concrete samples for decommission of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zapata-Garcia, Daniel; Wershofen, Herbert [Physikalisch-Technische Bundesanstalt (PTB), Bundesallee 100 38116, Braunschweig (Germany); Larijani, Cyrus; Sobrino-Petrirena, Maitane; Garcia-Miranda, Maria; Jerome, Simon M. [National Physical Laboratory (NPL), Hampton Road, Teddington, Middlesex, TW11 0LW (United Kingdom)

    2014-07-01

    Decommissioning of the oldest nuclear power reactors are some of the most challenging technological legacy issues many countries will face in forthcoming years, as many power reactors reach the end of their design lives. Decommissioning of nuclear reactors generates large amounts of waste that need to be classified according to their radioactive content. Approximately 10 % of the contaminated material ends up in different repositories (depending on their level of contamination) while the rest is decontaminated, measured and released into the environment or sent for recycling. Such classification needs to be done accurately in order to ensure that both the personnel involved in decommissioning and the population at large are not needlessly exposed to radiation or radioactive material and to minimise the environmental impact of such work. However, too conservative classification strategies should not be applied, in order to make proper use of radioactive waste repositories since space is limited and the full process must be cost-effective. Implicit in decommissioning and classification of waste is the need to analyse large amounts of material which usually combine a complex matrix with a non-homogeneous distribution of the radionuclides. Because the costs involved are large, it is possible to make great savings by the adoption of best available practices, such as the use of validated methods for on-site measurements and simultaneous determination of more than one radionuclide whenever possible. The work we present deals with the development and the validation of a procedure for the simultaneous determination of {sup 241}Am, plutonium isotopes, uranium isotopes and {sup 90}Sr in concrete samples. Samples are firstly ground and fused with LiBO{sub 2} and Li{sub 2}B{sub 4}O{sub 7}. After dissolution of the fused sample, silicate and alkaline elements are removed followed by radiochemical separation of the target radionuclides using extraction chromatography. Measurement

  14. Reactor-based management of used nuclear fuel: assessment of major options.

    Science.gov (United States)

    Finck, Phillip J; Wigeland, Roald A; Hill, Robert N

    2011-01-01

    This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms).

  15. Integration of advanced nuclear materials separation processes

    Energy Technology Data Exchange (ETDEWEB)

    Jarvinen, G.D.; Worl, L.A.; Padilla, D.D.; Berg, J.M.; Neu, M.P.; Reilly, S.D.; Buelow, S.

    1998-12-31

    This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). This project has examined the fundamental chemistry of plutonium that affects the integration of hydrothermal technology into nuclear materials processing operations. Chemical reactions in high temperature water allow new avenues for waste treatment and radionuclide separation.Successful implementation of hydrothermal technology offers the potential to effective treat many types of radioactive waste, reduce the storage hazards and disposal costs, and minimize the generation of secondary waste streams. The focus has been on the chemistry of plutonium(VI) in solution with carbonate since these are expected to be important species in the effluent from hydrothermal oxidation of Pu-containing organic wastes. The authors investigated the structure, solubility, and stability of the key plutonium complexes. Installation and testing of flow and batch hydrothermal reactors in the Plutonium Facility was accomplished. Preliminary testing with Pu-contaminated organic solutions gave effluent solutions that readily met discard requirements. A new effort in FY 1998 will build on these promising initial results.

  16. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wali, W A; Hassan, K H; Cullen, J D; Al-Shamma' a, A I; Shaw, A; Wylie, S R, E-mail: w.wali@2009.ljmu.ac.uk [Built Environment and Sustainable Technologies Institute (BEST), School of the Built Environment, Faculty of Technology and Environment Liverpool John Moores University, Byrom Street, Liverpool L3 3AF (United Kingdom)

    2011-08-17

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  17. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    Science.gov (United States)

    Wali, W. A.; Hassan, K. H.; Cullen, J. D.; Al-Shamma'a, A. I.; Shaw, A.; Wylie, S. R.

    2011-08-01

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  18. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Renier, J.A.

    2002-04-17

    Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron. Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized

  19. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  20. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  1. Improvement of nuclear ship engineering simulation system. Hardware renewal and interface improvement of the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroki; Kyoya, Masahiko; Shimazaki, Junya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kano, Tadashi [KCS, Co., Mito, Ibaraki (Japan); Takahashi, Teruo [Energis, Co., Kobe, Hyogo (Japan)

    2001-10-01

    JAERI had carried out the design study about a lightweight and compact integral type reactor (an advanced marine reactor) with passive safety equipment as a power source for the future nuclear ships, and completed an engineering design. We have developed the simulator for the integral type reactor to confirm the design and operation performance and to utilize the study of automation of the reactor operation. The simulator can be used also for future research and development of a compact reactor. However, the improvement in a performance of hardware and a human machine interface of software of the simulator were needed for future research and development. Therefore, renewal of hardware and improvement of software have been conducted. The operability of the integral-reactor simulator has been improved. Furthermore, this improvement with the hardware and software on the market brought about better versatility, maintainability, extendibility and transfer of the system. This report mainly focuses on contents of the enhancement in a human machine interface, and describes hardware renewal and the interface improvement of the integral type reactor simulator. (author)

  2. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  3. Work Domain Analysis Methodology for Development of Operational Concepts for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This report describes a methodology to conduct a Work Domain Analysis in preparation for the development of operational concepts for new plants. This method has been adapted from the classical method described in the literature in order to better deal with the uncertainty and incomplete information typical of first-of-a-kind designs. The report outlines the strategy for undertaking a Work Domain Analysis of a new nuclear power plant and the methods to be used in the development of the various phases of the analysis. Basic principles are described to the extent necessary to explain why and how the classical method was adapted to make it suitable as a tool for the preparation of operational concepts for a new nuclear power plant. Practical examples are provided of the systematic application of the method and the various presentation formats in the operational analysis of advanced reactors.

  4. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    Science.gov (United States)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  5. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  6. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW{sub th} and an electricity generation mode of 100 kW{sub th}, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  7. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Directory of Open Access Journals (Sweden)

    Seung Hyun Nam

    2015-10-01

    Full Text Available Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER, for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MWth and an electricity generation mode of 100 kWth, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  8. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  9. Review and Assessment of Neutron Cross Section and Nubar Covariances for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Maslov,V.M.; Oblozinsky, P.; Herman, M.

    2008-12-01

    In January 2007, the National Nuclear Data Center (NNDC) produced a set of preliminary neutron covariance data for the international project 'Nuclear Data Needs for Advanced Reactor Systems'. The project was sponsored by the OECD Nuclear Energy Agency (NEA), Paris, under the Subgroup 26 of the International Working Party on Evaluation Cooperation (WPEC). These preliminary covariances are described in two recent BNL reports. The NNDC used a simplified version of the method developed by BNL and LANL that combines the recent Atlas of Neutron Resonances, the nuclear reaction model code EMPIRE and the Bayesian code KALMAN with the experimental data used as guidance. There are numerous issues involved in these estimates of covariances and it was decided to perform an independent review and assessment of these results so that better covariances can be produced for the revised version in future. Reviewed and assessed are uncertainties for fission, capture, elastic scattering, inelastic scattering and (n,2n) cross sections as well as prompt nubars for 15 minor actinides ({sup 233,234,236}U, {sup 237}Np, {sup 238,240,241,242}Pu, {sup 241,242m,243}Am and {sup 242,243,244,245}Cm) and 4 major actinides ({sup 232}Th, {sup 235,238}U and {sup 239}Pu). We examined available evaluations, performed comparison with experimental data, taken into account uncertainties in model parameterization and made use state-of-the-art nuclear reaction theory to produce the uncertainty assessment.

  10. Above-ground antineutrino detection for nuclear reactor monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Sweany, M.; Brennan, J.; Cabrera-Palmer, B.; Kiff, S.; Reyna, D.; Throckmorton, D.

    2015-01-01

    Antineutrino monitoring of nuclear reactors has been demonstrated many times (Klimov et al., 1994 [1]; Bowden et al., 2009 [2]; Oguri et al., 2014 [3]), however the technique has not as of yet been developed into a useful capability for treaty verification purposes. The most notable drawback is the current requirement that detectors be deployed underground, with at least several meters-water-equivalent of shielding from cosmic radiation. In addition, the deployment of liquid-based detection media presents a challenge in reactor facilities. We are currently developing a detector system that has the potential to operate above ground and circumvent deployment problems associated with a liquid detection media: the system is composed of segments of plastic scintillator surrounded by {sup 6}LiF/ZnS:Ag. ZnS:Ag is a radio-luminescent phosphor used to detect the neutron capture products of {sup 6}Li. Because of its long decay time compared to standard plastic scintillators, pulse-shape discrimination can be used to distinguish positron and neutron interactions resulting from the inverse beta decay (IBD) of antineutrinos within the detector volume, reducing both accidental and correlated backgrounds. Segmentation further reduces backgrounds by identifying the positron's annihilation gammas, a signature that is absent for most correlated and uncorrelated backgrounds. This work explores different configurations in order to maximize the size of the detector segments without reducing the intrinsic neutron detection efficiency. We believe that this technology will ultimately be applicable to potential safeguards scenarios such as those recently described by Huber et al. (2014) [4,5].

  11. Controlling the power output of a nuclear reactor with fuzzy logic

    NARCIS (Netherlands)

    Ruan, D.; Wal, A.J. van der

    1998-01-01

    The application of fuzzy logic control (FLC) in the domain of nuclear industry presents a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear reactors and the very strict safety regulations in force for nuclear power plants. The very same regulations preven

  12. Controlling the Power Output of a Nuclear Reactor with Fuzzy Logic

    NARCIS (Netherlands)

    Ruan, D.; Wal, A.J. van der

    1997-01-01

    The application of fuzzy logic control (FLC) in the domain of nuclear industry presents a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear reactors and the very strict safety regulations in force for nuclear power plants. The very same regulations preven

  13. The near boiling reactor : conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Energy Technology Data Exchange (ETDEWEB)

    Cole, C.J.P

    2005-07-01

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the 'Victoria' Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96{sup o}C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has

  14. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Science.gov (United States)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achi