WorldWideScience

Sample records for oxide reactor program

  1. Neutron study of fast neutron reactor systems by exponential experiments on Harmonie - Graphite program HUG-PHUG - Oxide program PHRIXOS - Uranium program UK

    International Nuclear Information System (INIS)

    Desprets, Alain.

    1977-12-01

    Exponential experiments allow to obtain the fundamental characteristics of a lattice (material buckling, reaction rate ratios) more economically than critical experiments. This report describes the experimental techniques and the methods of analysis used for this type of experiments. The results obtained with three programs performed with the source reactor HARMONIE are given: graphite-lattices program (3 U-fueled and 3 Pu-fueled lattices); oxide-fuel program (4 PuO 2 -UO 2 lattices); pure uranium program (one lattice). Some of these lattices were also studied in critical experiments. The coherence of the results obtained by the two types of experiments is established [fr

  2. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  3. Quality assurance program for surveillance of fast reactor mixed oxide fuel analytical chemistry

    International Nuclear Information System (INIS)

    Rein, J.E.; Zeigler, R.K.; Waterbury, G.R.; McClung, W.E.; Praetorius, P.R.; Delvin, W.L.

    1976-01-01

    An effective quality assurance program for the chemical analysis of nuclear fuel is essential to assure that the fuel will meet the strict chemical specifications required for optimum reactor performance. Such a program has been in operation since 1972 for the fuels manufactured for the Fast Flux Test Facility. This program, through the use of common quality control and calibration standards, has consistently provided high levels of agreement among laboratories in all areas of analysis. The paper presented gives a summary of the chemical specifications for the fuel and source material, an outline of the requirements for laboratory qualifications and the preparation of calibration and quality control materials, general administration details of the plan, and examples where the program has been useful in solving laboratory problems

  4. Program of quality management when fabricating fast reactor vibropack oxide fuel pins

    International Nuclear Information System (INIS)

    Mayorshin, A.A.; Kisly, V.A.; Sudakov, L.V.

    2000-01-01

    There are presented main principles of creation and operation of Quality Management Program in fabricating vibropack oxide fuel pins for BOR-60 and BN-600 being in force in SSC RF RIAR. There is given structure of documentation for QS principal elements. Under Quality System there are defined all the procedures, assuring that fuel pin meets the normative requirements. The system model is complied with the standard model IS 9001. There are shown technologic flowchart and check operation, statistic results of pin critical parameter check as well as main results of in-pile tests. (author)

  5. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Tehan, Terry

    2002-01-01

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  6. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  7. University Reactor Instrumentation Program

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1992-11-01

    Recognizing that the University Reactor Instrumentation Program was developed in response to widespread needs in the academic community for modernization and improvement of research and training reactors at institutions such as the University of Florida, the items proposed to be supported by this grant over its two year period have been selected as those most likely to reduce foreed outages, to meet regulatory concerns that had been expressed in recent years by Nuclear Regulatory Commission inspectors or to correct other facility problems and limitations. Department of Energy Grant Number DE-FG07-90ER129969 was provided to the University of Florida Training Reactor(UFTR) facility through the US Department of Energy's University Reactor Instrumentation Program. The original proposal submitted in February, 1990 requested support for UFTR facility instrumentation and equipment upgrades for seven items in the amount of $107,530 with $13,800 of this amount to be the subject of cost sharing by the University of Florida and $93,730 requested as support from the Department of Energy. A breakdown of the items requested and total cost for the proposed UFTR facility instrumentation and equipment improvements is presented

  8. Oxidizer in phosphoric reactors

    International Nuclear Information System (INIS)

    Santos Benedetto, J. dos

    1985-01-01

    Oxidation during the manufacture of wet-process phosphoric acid affected the distribution of uranium and impurities between phosphoric acid and gypsum, by decreasing the uranium loss to gypsum and the impurities solubilization in phosphoric acid. (Author) [pt

  9. Nuclear Reactor Sharing Program

    International Nuclear Information System (INIS)

    1994-01-01

    The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

  10. 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.

  11. University Reactor Sharing Program

    International Nuclear Information System (INIS)

    Reese, W.D.

    2004-01-01

    Research projects supported by the program include items such as dating geological material and producing high current super conducting magnets. The funding continues to give small colleges and universities the valuable opportunity to use the NSC for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy. The Reactor Sharing Program has supported the construction of a Fast Neutron Flux Irradiator for users at New Mexico Institute of Mining and Technology and the University of Houston. This device has been characterized and has been found to have near optimum neutron fluxes for A39/Ar 40 dating. Institution final reports and publications resulting from the use of these funds are on file at the Nuclear Science Center

  12. The US Liquid Metal Reactor Development Program

    International Nuclear Information System (INIS)

    Till, C.E.; Arnold, W.H.; Griffith, J.D.

    1988-01-01

    The US Liquid Metal Reactor Development Program has been restructured to take advantage of the opportunity today to carry out R and D on truly advanced reactor technology. The program gives particular emphasis to improvements to reactor safety. The new directions are based on the technology of the Integral Fast Reactor (IFR). Much of the basis for superior safety performance using IFR technology has been experimentally verified and aggressive programs continue in EBR-II and TREAT. Progress has been made in demonstrating both the metallic fuel and the new electrochemical processes of the IFR. The FFTF facility is converting to metallic fuel; however, FFTF also maintains a considerable US program in oxide fuels. In addition, generic programs are continuing in steam generator testing, materials development, and, with international cooperation, aqueous reprocessing. Design studies are carried out in conjunction with the IFR technology development program. In summary, the US maintains an active development program in Liquid Metal Reactor technology, and new directions in reactor safety are central to the program

  13. Solid oxide electrochemical reactor science.

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, Neal P. (Colorado School of Mines, Golden, CO); Stechel, Ellen Beth; Moyer, Connor J. (Colorado School of Mines, Golden, CO); Ambrosini, Andrea; Key, Robert J. (Colorado School of Mines, Golden, CO)

    2010-09-01

    Solid-oxide electrochemical cells are an exciting new technology. Development of solid-oxide cells (SOCs) has advanced considerable in recent years and continues to progress rapidly. This thesis studies several aspects of SOCs and contributes useful information to their continued development. This LDRD involved a collaboration between Sandia and the Colorado School of Mines (CSM) ins solid-oxide electrochemical reactors targeted at solid oxide electrolyzer cells (SOEC), which are the reverse of solid-oxide fuel cells (SOFC). SOECs complement Sandia's efforts in thermochemical production of alternative fuels. An SOEC technology would co-electrolyze carbon dioxide (CO{sub 2}) with steam at temperatures around 800 C to form synthesis gas (H{sub 2} and CO), which forms the building blocks for a petrochemical substitutes that can be used to power vehicles or in distributed energy platforms. The effort described here concentrates on research concerning catalytic chemistry, charge-transfer chemistry, and optimal cell-architecture. technical scope included computational modeling, materials development, and experimental evaluation. The project engaged the Colorado Fuel Cell Center at CSM through the support of a graduate student (Connor Moyer) at CSM and his advisors (Profs. Robert Kee and Neal Sullivan) in collaboration with Sandia.

  14. N Reactor Deactivation Program Plan

    International Nuclear Information System (INIS)

    Walsh, J.L.

    1993-12-01

    This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities · in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directive to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually

  15. Breazeale Reactor Modernization Program

    International Nuclear Information System (INIS)

    Davison, C. C.

    2003-01-01

    The Penn State Breazeale Nuclear Reactor is the longest operating licensed research reactor in the nation. The facility has played a key role in educating scientists, engineers and in providing facilities and services to researchers in many different disciplines. In order to remain a viable and effective research and educational institution, a multi-phase modernization project was proposed. Phase I was the replacement of the 25-year old reactor control and safety system along with associated wiring and hardware. This phase was fully funded by non-federal funds. Tasks identified in Phases II-V expand upon and complement the work done in Phase I to strategically implement state-of-the-art technologies focusing on identified national needs and priorities of the future

  16. The 'SURA' fast reactor program

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The Commissariat a l'Energie Atomique's SURA program on fast reactor safety consists of two specific testing programs on fastbreeder reactor safety: the Cabri and Scarabee programs. Both Cabri and Scarabee are examples of multinational research collaboration. The CEA and the Karlsruhe Nuclear Research Center are each covering half of the construction costs. Britain, the US and Japan are also due to participate in these experiments. The aim of the programs is to examine the behaviour of fuel in sodium cooled fast reactors. The Cabri program consists of setting off a reactivity accident in a power reactor core which is cooled with liquid sodium, such an accident occurring after a sharp increase in reactivity or as a result of the pump suddenly breaking down without there at the same time being any fall in the control rods. In 1967 the Commissariat a l'Energie Atomique started its Scarabee research program which is trying to analyse the sort of things that can go wrong with fuel cooling systems and what the consequences can be [fr

  17. New Production Reactors Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-01

    Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs.

  18. New Production Reactors Program Plan

    International Nuclear Information System (INIS)

    1990-12-01

    Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs

  19. Nuclear reactor shield including magnesium oxide

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1981-01-01

    An improvement is described for nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux. The reactor shielding includes means providing structural support, neutron moderator material, neutron absorber material and other components, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron

  20. University Reactor Matching Grants Program

    International Nuclear Information System (INIS)

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-01-01

    During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given

  1. Oxidative coupling of methane using inorganic membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Y.H.; Moser, W.R.; Dixon, A.G. [Worcester Polytechnic Institute, MA (United States)] [and others

    1995-12-31

    The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gas phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.

  2. Staged membrane oxidation reactor system

    Science.gov (United States)

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2012-09-11

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  3. Plutonium bearing oxide fuels for recycling in thermal reactors and fast breeder reactors

    International Nuclear Information System (INIS)

    Cunningham, G.W.

    1977-01-01

    Programs carried out in the past two decades have established the technical feasibility of using plutonium as a fuel material in both water-cooled power reactors and sodium-cooled fast breeder reactors. The problem facing the technical community is basically one of demonstrating plutonium fuel recycle under strict conditions of public safety, accountability, personnel exposure, waste management, transportation and diversion or theft which are still evolving. In this paper only technical and economic aspects of high volume production and the demonstration program required are discussed. This paper discusses the role of mixed oxide fuels in light water reactors and the objectives of the LMFBR required for continual growth of nuclear power during the next century. The results of studies showing the impact of using plutonium on uranium requirements, power costs, and the market share of nuclear power are presented. The influence of doubling time and the introduction date of LMFBRs on the benefits to be derived by its commercial use are discussed. Advanced fuel development programs scoped to meet future commerical LMFBR fuel requirements are described. Programs designed to provide the basic technology required for using plutonium fuels in a manner which will satisfy all requirements for public acceptance are described. Included are the high exposure plutonium fabrication development program centered around the High Performance Fuels Laboratory being built at the Hanford Engineering Development Laboratory and the program to confirm the technology required for the production of mixed oxide fuels for light water reactors which is being coordinated by Savannah River Laboratories

  4. Advances by the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs

  5. The United States Advanced Reactor Technologies Research and Development Program

    International Nuclear Information System (INIS)

    O’Connor, Thomas J.

    2014-01-01

    The following aspects are addressed: • Nuclear energy mission; • Reactor research development and deployment (RD&D) programs: - Light Water Reactor Sustainability Program; - Small Modular Reactor Licensing Technical Support; - Advanced Reactor Technologies (ART)

  6. NCSU reactor sharing program. Final technical report

    International Nuclear Information System (INIS)

    Perez, P.B.

    1997-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996

  7. U.S. Domestic Reactor Conversion Programs

    International Nuclear Information System (INIS)

    Woolstenhulme, Eric

    2008-01-01

    The Conversion Projects Include: the revision of the facilities safety basis documents and supporting analysis, the fabrication of new LEU fuel, the change-out of the reactor core, and the removal of the used HEU fuel (by INL University Fuels Program or DOE-NE). The major entities involved are: the U.S. Nuclear Regulatory Commission, the University reactor department, the fuel and hardware fabricators, the Spent fuel receipt facilities, the Spent fuel shipping services, and the U.S. Department of Energy and their subcontractors. Three major Reactor Conversion Program milestones have been accomplished since 2006: the conversion of the TRIGA reactor at Texas A and M University Nuclear Science Center, the conversion of the University of Florida Training Reactor, and the conversion of the Purdue University Reactor. Four Reactor Conversion Program milestones yet to be accomplished in 2008 and 2009: the Washington State University Nuclear Radiation Center reactor, the Oregon State University TRIGA Reactor, the University of Wisconsin Nuclear Reactor, and the Neutron Radiography Reactor Facility. NNSA is committed to doing things cheaper, better, smarter, safer through a 'Lessons Learned' process. The conversion team assessed each major activity grouping: Project Initiation, Conversion Proposal Development, Fuel Fabrication and Hardware, Core Conversion, and Spent Nuclear Fuel Removal. Issues were identified and recommendations were given

  8. Programming for a nuclear reactor instrument simulator

    International Nuclear Information System (INIS)

    Cohn, C.E.

    1989-01-01

    A new computerized control system for a transient test reactor incorporates a simulator for pre-operational testing of control programs. The part of the simulator pertinent to the discussion here consists of two microprocessors. An 8086/8087 reactor simulator calculates simulated reactor power by solving the reactor kinetics equations. An 8086 instrument simulator takes the most recent power value developed by the reactor simulator and simulates the appropriate reading on each of the eleven reactor instruments. Since the system is required to run on a one millisecond cycle, careful programming was required to take care of all eleven instruments in that short time. This note describes the special programming techniques used to attain the needed performance

  9. Reactor use in nuclear engineering programs

    International Nuclear Information System (INIS)

    Murray, R.L.

    1975-01-01

    Nuclear reactors for dual use in training and research were established at about 50 universities in the period since 1950, with assistance by the U. S. Atomic Energy Commission and the National Science Foundation. Most of the reactors are in active use for a variety of educational functions--laboratory teaching of undergraduates and graduate students, graduate research, orientation of visitors, and nuclear power plant reactor operator training, along with service to the technical community. As expected, the higher power reactors enjoy a larger average weekly use. Among special programs are reactor sharing and high-school teachers' workshops

  10. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  11. The US Liquid-Metal Reactor Program - overview and status

    International Nuclear Information System (INIS)

    Quinn, J.E.; Gyorey, G.L.; Salerno, L.N.

    1992-01-01

    The US Advanced Liquid-Metal Reactor (ALMR) Program has three major elements being developed in an integrated fashion to produce a system meeting the US long-term nuclear energy needs. Reactor design, one of those elements, is the focus of this paper. The other two elements, the integral fast reactor metal-fuel cycle and the light water reactor (LWR) spent-fuel actinide recycle, will be addressed in companion papers. The ALMR is adaptable to multiple missions with few modifications such as the core arrangements. The missions identified to date are (a) the extension of the existing uranium resources through breeding and highly efficient uranium utilization, (b) the recycle and utilization of the long-life actinides in LWR spent fuel as fissile material for the ALMR, and (c) the conversion of excess weapons fissil material into electricity. In addition to these missions, the reactor design is adaptable to either the metal-fuel cycle or the oxide fuel cycle

  12. Impacts on power reactor health physics programs

    International Nuclear Information System (INIS)

    Meyer, B.A.

    1991-01-01

    The impacts on power reactor health physics programs form implementing the revised 10 CFR Part 20 will be extensive and costly. Every policy, program, procedure and training lesson plan involving health physics will require changes and the subsequent retraining of personnel. At each power reactor facility, hundreds of procedures and thousands of people will be affected by these changes. Every area of a power reactor health physics program will be affected. These areas include; ALARA, Respiratory Protection, Exposure Control, Job Coverage, Dosimetry, Radwaste, Effluent Accountability, Emergency Planning and Radiation Worker Training. This paper presents how power reactor facilities will go about making these changes and gives possible examples of some of these changes and their impact on each area of power reactor health physics program

  13. The experimental program of neutronphysics for advanced water reactors

    International Nuclear Information System (INIS)

    Martin-Deider, L.; Cathalu, S.; Santamarina, A.; Gomit, M.

    1985-11-01

    The C.E.A. and E.D.F. has jointly undertaken a program of experimental studies on under-moderated water lattices, with mixed oxide fuel UO 2 -PuO 2 . Undermoderated lattices offer high conversion ratios. This type of lattice could limit in the future the natural uranium consumption of pressurized water reactors. This experimental program is aimed at qualifying neutron transport calculations in a large range of moderating ratio (between 0.5 and 1.5). It includes three experiments: ERASME, a critical experiment of large size in the EOLE reactor at Cadarache; ICARE, an irradiation experiment in the MELUSINE reactor at Grenoble; and an experiment to measure the reactivity effects by oscillations in the MINERVE reactor at Cadarache [fr

  14. The advanced test reactor strategic evaluation program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1989-01-01

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed

  15. Catalytic Reactor For Oxidizing Mercury Vapor

    Science.gov (United States)

    Helfritch, Dennis J.

    1998-07-28

    A catalytic reactor (10) for oxidizing elemental mercury contained in flue gas is provided. The catalyst reactor (10) comprises within a flue gas conduit a perforated corona discharge plate (30a, b) having a plurality of through openings (33) and a plurality of projecting corona discharge electrodes (31); a perforated electrode plate (40a, b, c) having a plurality of through openings (43) axially aligned with the through openings (33) of the perforated corona discharge plate (30a, b) displaced from and opposing the tips of the corona discharge electrodes (31); and a catalyst member (60a, b, c, d) overlaying that face of the perforated electrode plate (40a, b, c) opposing the tips of the corona discharge electrodes (31). A uniformly distributed corona discharge plasma (1000) is intermittently generated between the plurality of corona discharge electrode tips (31) and the catalyst member (60a, b, c, d) when a stream of flue gas is passed through the conduit. During those periods when corona discharge (1000) is not being generated, the catalyst molecules of the catalyst member (60a, b, c, d) adsorb mercury vapor contained in the passing flue gas. During those periods when corona discharge (1000) is being generated, ions and active radicals contained in the generated corona discharge plasma (1000) desorb the mercury from the catalyst molecules of the catalyst member (60a, b, c, d), oxidizing the mercury in virtually simultaneous manner. The desorption process regenerates and activates the catalyst member molecules.

  16. Light water reactor mixed-oxide fuel irradiation experiment

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-01-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

  17. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  18. Reactor Containment Spray Technology Program

    Energy Technology Data Exchange (ETDEWEB)

    Row, T. H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    1968-12-15

    The design basis accident in water moderated power reactors is a loss-of-coolant accident in which water sprays are generally employed to control the containment pressure transient by condensing the released steam-air mixture. Additives to the spray have been proposed as a way to increase their usefulness by enhancing the removal of various forms of radioiodine from the containment atmosphere. A program to investigate the gas-liquid systems involved is co-ordinated by ORNL for the US Atomic Energy Commission. A basic part of the program is the search for various chemical additives that will increase the spray affinity for molecular iodine and methyl iodide. A method for evaluating additives was developed that measures equilibrium distribution coefficients for iodine between air and aqueous solutions. Additives selected are used in single drop-wind tunnel experiments where the circulating gas contains iodine or CH{sub 3}I. Mass transfer coefficients and transient distribution coefficients have been determined as a function of relative humidity, temperature, drop size, and solution pH and concentration. Tests have shown that surfactants and organic amines increase the solution ability to getter CH{sub 3}l. Results from single drop tests help in planning spray experiments in the Nuclear Safety Pilot Plant, a large ({approx}38 m{sup 3}) facility, where accident conditions are closely simulated. Iodine and CH{sub 3}I removal rates have been determined for a number of solutions, including 1 wt% Na{sub 2}S{sub 2}O{sub 3} + 3000 ppm B + 0.153 M NaOH and 3000 ppm B + 0.153 M NaOH. The additive has very little effect in removal of I{sub 2} with half-lives of less than 1 mm typical for any aqueous solution. These same solutions remove CH{sub 3}I with a half-life of one hour. Analytical models for the removal processes have been developed. Consideration is also being given to corrosion, thermal and radiation stability of the solutions. Radiation studies have indicated the loss

  19. Nuclear Power Reactor simulator - based training program

    International Nuclear Information System (INIS)

    Abdelwahab, S.A.S.

    2009-01-01

    nuclear power stations will continue playing a major role as an energy source for electric generation and heat production in the world. in this paper, a nuclear power reactor simulator- based training program will be presented . this program is designed to aid in training of the reactor operators about the principles of operation of the plant. also it could help the researchers and the designers to analyze and to estimate the performance of the nuclear reactors and facilitate further studies for selection of the proper controller and its optimization process as it is difficult and time consuming to do all experiments in the real nuclear environment.this program is written in MATLAB code as MATLAB software provides sophisticated tools comparable to those in other software such as visual basic for the creation of graphical user interface (GUI). moreover MATLAB is available for all major operating systems. the used SIMULINK reactor model for the nuclear reactor can be used to model different types by adopting appropriate parameters. the model of each component of the reactor is based on physical laws rather than the use of look up tables or curve fitting.this simulation based training program will improve acquisition and retention knowledge also trainee will learn faster and will have better attitude

  20. Pathfinder irradiation of advanced fuel (Th/U mixed oxide) in a power reactor

    International Nuclear Information System (INIS)

    Brant Pinheiro, R.

    1993-01-01

    Within the joint Brazilian-German cooperative R and D Program on Thorium Utilization in Pressurized Water Reactors carried out from 1979 to 1988 by Nuclebras/CDTN, KFA-Juelich, Siemens/KWU and NUKEM, a pathfinder irradiation of Th/U mixed oxide fuel in the Angra 1 nuclear power reactor was planned. The objectives of this irradiation testing, the irradiation strategy, the work performed and the status achieved at the end of the joint Program are presented. (author)

  1. Program summary for the Civilian Reactor Development Program

    International Nuclear Information System (INIS)

    1982-07-01

    This Civilian Reactor Development Program document has the prime purpose of summarizing the technical programs supported by the FY 1983 budget request. This section provides a statement of the overall program objectives and a general program overview. Section II presents the technical programs in a format intended to show logical technical interrelationships, and does not necessarily follow the structure of the formal budget presentation. Section III presents the technical organization and management structure of the program

  2. Developmental Light-Water Reactor Program

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-12-01

    This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

  3. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    Moore, Mark; George, Robert; Spence, Harry; Nguyen, John

    1992-01-01

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  4. IRSN research programs concerning reactor safety

    International Nuclear Information System (INIS)

    Bardelay, J.

    2005-01-01

    This paper is made up of 3 parts. The first part briefly presents the missions of IRSN (French research institute on nuclear safety), the second part reviews the research works currently led by IRSN in the following fields : -) the assessment of safety computer codes, -) thermohydraulics, -) reactor ageing, -) reactivity accidents, -) loss of coolant, -) reactor pool dewatering, -) core meltdown, -) vapor explosion, and -) fission product release. In the third part, IRSN is shown to give a major importance to experimental programs led on research or test reactors for collecting valid data because of the complexity of the physical processes that are involved. IRSN plans to develop a research program concerning the safety of high or very high temperature reactors. (A.C.)

  5. Trends in light water reactor dosimetry programs

    International Nuclear Information System (INIS)

    Rahn, F.J.; Serpan, C.Z.; Fabry, A.; McElroy, W.N.; Grundl, J.A.; Debrue, J.

    1977-01-01

    Dosimetry programs and techniques play an essential role in the continued assurance of the safety and reliability of components of light water reactors. Primary concern focuses on the neutron irradiation embrittlement of reactor pressure vessels and methods by which the integrity of a pressure vessel can be predicted and monitored throughout its service life. Research in these areas requires a closely coordinated program which integrates the elements of the calculational and material sciences, the development of advanced dosimetric techniques and the use of benchmarks and validation of these methods. The paper reviews the status of the various international efforts in the dosimetry area

  6. The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

    2001-07-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

  7. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Calabrese, Carlos R.

    2001-01-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  8. Programming for a nuclear reactor instrument simulation

    International Nuclear Information System (INIS)

    Cohn, C.

    1988-01-01

    This note discusses 8086/8087 machine-language programming for simulation of nuclear reactor instrument current inputs by means of a digital-analog converter (DAC) feeding a bank of series input resistors. It also shows FORTRAN programming for generating the parameter tales used in the simulation. These techniques would be generally useful for high-speed simulation of quantities varying over many orders of magnitude

  9. Operation of staged membrane oxidation reactor systems

    Science.gov (United States)

    Repasky, John Michael

    2012-10-16

    A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

  10. EPRI program in water reactor safety

    International Nuclear Information System (INIS)

    Loewenstein, W.B.; Gelhaus, F.; Gopalakrishnan, A.

    1975-01-01

    The basis for EPRI's water reactor safety program is twofold. First is compilation and development of fundamental background data necessary for quantified light-water reactor (LWR) safety assurance appraisals. Second is development of realistic and experimentally bench-marked analytical procedures. The results are expected either to confirm the safety margins in current operating parameters, and to identify overly conservative controls, or, in some cases, to provide a basis for improvements to further minimize uncertainties in expected performance. Achievement of these objectives requires the synthesis of related current and projected experimental-analytical projects toward establishment of an experimentally-based analysis for the assurance of safety for LWRs

  11. Flow analysis in a supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Oh, C.H.; Kochan, R.J.; Beller, J.M.

    1996-01-01

    Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency

  12. Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

    Science.gov (United States)

    Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

    1994-01-01

    This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

  13. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1999-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  14. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  15. Engineering Design of a Double Reactor for Spent Fuel Oxidation

    International Nuclear Information System (INIS)

    Kim, Young-Hwan; Lee, Jae-Won; Lee, Ju-Ho; Cho, Yung-Zun; Ahn, Do-Hee

    2015-01-01

    In this study, for a performance enhancement of the oxidation treatment device recovery ratio, the first performance test of the existing device (prototype) oxidation treatment device was carried out. In addition, by analyzing the result, the size of the reactor with a 1 kg HM/batch for a recovery ratio enhancement was decided, and the structure of the reactor was derived as a double structure reactor with a mesh type drum. The principle and structure of this device are as follows. The pellet of the supplied rods is oxidized in 500 .deg. C reactor A, and penetrates reactor B to form a uniform powder. In addition, if it is rotated in the reverse direction, the powder and hull are separated. The device is composed of a reactor module, driving module, heater module, support module, outlet module, etc. In addition, by reflecting the enhancements, a voloxidizer with a double reactor was designed and manufactured, and a second performance test was carried out. Using a 30 mm hull and simulated powders (balls), as a result of carrying out the enhanced device performance test, the hull recovery ratio was 100%, and the simulated powder recovery ratio was 99% or more

  16. Engineering Design of a Double Reactor for Spent Fuel Oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young-Hwan; Lee, Jae-Won; Lee, Ju-Ho; Cho, Yung-Zun; Ahn, Do-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this study, for a performance enhancement of the oxidation treatment device recovery ratio, the first performance test of the existing device (prototype) oxidation treatment device was carried out. In addition, by analyzing the result, the size of the reactor with a 1 kg HM/batch for a recovery ratio enhancement was decided, and the structure of the reactor was derived as a double structure reactor with a mesh type drum. The principle and structure of this device are as follows. The pellet of the supplied rods is oxidized in 500 .deg. C reactor A, and penetrates reactor B to form a uniform powder. In addition, if it is rotated in the reverse direction, the powder and hull are separated. The device is composed of a reactor module, driving module, heater module, support module, outlet module, etc. In addition, by reflecting the enhancements, a voloxidizer with a double reactor was designed and manufactured, and a second performance test was carried out. Using a 30 mm hull and simulated powders (balls), as a result of carrying out the enhanced device performance test, the hull recovery ratio was 100%, and the simulated powder recovery ratio was 99% or more.

  17. OXIDATIVE COUPLING OF METHANE USING INORGANIC MEMBRANE REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Y.H. Ma; Dr. W.R. Moser; Dr. A.G. Dixon; Dr. A.M. Ramachandra; Dr. Y. Lu; C. Binkerd

    1998-04-01

    The objective of this research is to study the oxidative coupling of methane in catalytic inorganic membrane reactors. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and higher yields than in conventional non-porous, co-feed, fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gas phase reactions, which are believed to be a main route for the formation of CO{sub x} products. Such gas phase reactions are a cause of decreased selectivity in the oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Membrane reactor technology also offers the potential for modifying the membranes both to improve catalytic properties as well as to regulate the rate of the permeation/diffusion of reactants through the membrane to minimize by-product generation. Other benefits also exist with membrane reactors, such as the mitigation of thermal hot-spots for highly exothermic reactions such as the oxidative coupling of methane. The application of catalytically active inorganic membranes has potential for drastically increasing the yield of reactions which are currently limited by either thermodynamic equilibria, product inhibition, or kinetic selectivity.

  18. Oxidation during reflood of reactor core with melting cladding

    Energy Technology Data Exchange (ETDEWEB)

    Siefken, L.J.; Allison, C.M.; Davis, K.L. [and others

    1995-09-01

    Models were recently developed and incorporated into the SCDAP/RELAP5 code for calculating the oxidation of fuel rods during cladding meltdown and reflood. Experiments have shown that a period of intense oxidation may occur when a hot partially oxidized reactor core is reflooded. This paper offers an explanation of the cladding meltdown and oxidation processes that cause this intense period of oxidation. Models for the cladding meltdown and oxidation processes are developed. The models are assessed by simulating a severe fuel damage experiment that involved reflood. The models for cladding meltdown and oxidation were found to improve calculation of the temperature and oxidation of fuel rods during the period in which hot fuel rods are reflooded.

  19. The Advanced Test Reactor Strategic Evaluation Program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1990-01-01

    A systematic evaluation of safety, environmental, and operational issues has been initiated at the Advanced Test Reactor (ATR). This program, the Strategic Evaluation Program (STEP), provides an integrated review of safety and operational issues against the standards applied to licensed commercial facilities. In the review of safety issues, 18 deviations were identified which required prompt attention. Resolution of these items has been accelerated in the program. An integrated living schedule is being developed to address the remaining findings. A risk evaluation is being performed on the proposed corrective actions and these actions will then be formally ranked in order of priority based on considerations of safety and operational significance. Once the final ranking is completed, an integrated schedule will be developed, which will include considerations of availability of funding and operating schedule. 3 refs., 2 figs

  20. Process for producing nuclear reactor fuel oxides

    International Nuclear Information System (INIS)

    Goenrich, H.; Druckenbrodt, W.G.

    1981-01-01

    The waste gases of the calcination process furnace in the AVC or AV/PuC process (manufacture of nuclear reactor fuel dioxides) are returned to the furnace in a closed circuit. The NH 3 produced replaces the hydrogen which would otherwise be required for reduction in this process. (orig.) [de

  1. Program of RA reactor start-up to nominal power

    International Nuclear Information System (INIS)

    1959-01-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is described in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members

  2. Manufacture of components for Canadian reactor programs

    International Nuclear Information System (INIS)

    Perry, L.P.

    Design features, especially those relating to calandrias, are pointed out for many CANDU-type reactors and the Taiwan research reactor. The special requirements shouldered by the Canadian suppliers of heavy reactor components are analyzed. (E.C.B.)

  3. Compatibility of sodium with ceramic oxides employed in nuclear reactors

    International Nuclear Information System (INIS)

    Acena, V.

    1981-01-01

    A review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors, is presented. The results of a thermo-dynamic/liquid sodium reactions are included. The exercise has been conducted with a view to effecting experimental studies in the future. (author) [es

  4. Compatibility of sodium with ceramic oxides employed in nuclear reactors

    International Nuclear Information System (INIS)

    Acena Moreno, V.

    1981-01-01

    This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

  5. Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

    KAUST Repository

    Bin Hamzah, Muhamad Firdaus

    2017-05-01

    Ignition and combustion characteristics of n-butanol/air, 2-butanol.air and isobutanol/air mixtures at stoichiometric (ϕ = 1) and lean (ϕ = 0.5) conditions were investigated in a micro flow reactor with a controlled temperature profile from 323 K to 1313 K, under atmospheric pressure. Sole distinctive weak flame was observed for each mixture, with inlet fuel/air mixture velocity set low at 2 cm/s. One-dimensional computation with comprehensive chemistry and transport was conducted. At low mixture velocities, one-stage oxidation was confirmed from heat release rate profiles, which was broadly in agreement with the experimental results. The weak flame positions were congruent with literature describing reactivity of the butanol isomers. These weak flame responses were also found to mirror the trend in Anti-Knock Indexes of the butanol isomers. Flux and sensitivity analyses were performed to investigate the fuel oxidation pathways at low and high temperatures. Further computational investigations on oxidation of butanol isomers at higher pressure of 5 atm indicated two-stage oxidation through the heat release rate profiles. Low temperature chemistry is accentuated in the region near the first weak cool flame for oxidation under higher pressure, and its impact on key species – such as hydroxyl radical, hydrogen peroxide and carbon monoxide – were considered. Both experimental and computational findings demonstrate the advantage of employing the micro flow reactor in investigating oxidation processes in the temperature region of interest along the reactor channel. By varying physical conditions such as pressure, the micro flow reactor system is proven to be highly beneficial in elucidating oxidation behavior of butanol isomers in conditions in engines such as those that mirror HCCI operations.

  6. Experience with oxide fuel for advanced reactors

    International Nuclear Information System (INIS)

    Leggett, R.D.

    1984-01-01

    This paper focuses on the use and potential of oxide fuel systems for the LMFBR. The flawless performance of mixed oxide (UO 2 -PuO 2 ) fuel in FFTF to 100,000 MWd/MTM is reviewed and means for achieving 200,000 MWd/MTM are presented. This includes using non-swelling alloys for cladding and ducts to overcome the limitations caused by swelling of the current alloys. Examples are provided of the inherently safe characteristics of oxide fuel including a large negative Doppler coefficient, its dispersive nature under hypothetical accident scenarios, and the low energy molten fuel-coolant interaction. Developments in fuel fabrication and reprocessing that stress safety and reduced personnel exposure are presented. Lastly, the flexibility to design for maximum fuel supply (high breeding gain) or minimum fuel cost (long lifetime) is shown

  7. Experience with oxide fuel for advanced reactors

    International Nuclear Information System (INIS)

    Leggett, R.D.

    1984-04-01

    This paper focuses on the use and potential of oxide fuel system for the LMFBR. The flawless performance of mixed oxide (UO 2 -PuO 2 ) fuel in FFTF to 100,000 MWd/MTM is reviewed and means for achieving 200,000 MWd/MTM are presented. This includes using non-swelling alloys for cladding and ducts to overcome the limitations caused by swelling of the current alloys. Exampled are provided of the inherently safe characteristics of oxide fuel including a large negative Doppler coefficient, its dispersive nature under hypothetical accident scenarios, and the low energy molten fuel-coolant interaction. Developments in fuel fabrication and reprocessing that stress safety and reduced personnel exposure are presented. Lastly, the flexibility to design for maximum fuel supply (high breeding gain) or minimum fuel cost (long lifetime) is shown

  8. Impact of Oxidative Stress in Fetal Programming

    Directory of Open Access Journals (Sweden)

    Loren P. Thompson

    2012-01-01

    Full Text Available Intrauterine stress induces increased risk of adult disease through fetal programming mechanisms. Oxidative stress can be generated by several conditions, such as, prenatal hypoxia, maternal under- and overnutrition, and excessive glucocorticoid exposure. The role of oxidant molecules as signaling factors in fetal programming via epigenetic mechanisms is discussed. By linking oxidative stress with dysregulation of specific target genes, we may be able to develop therapeutic strategies that protect against organ dysfunction in the programmed offspring.

  9. Programming Guidelines for FBD Programs in Reactor Protection System Software

    International Nuclear Information System (INIS)

    Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom; Lee, Jang Su

    2014-01-01

    Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows

  10. Programming Guidelines for FBD Programs in Reactor Protection System Software

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom [Division of Computer Science and Engineering College of Information and Communication, Konkuk University, Seoul (Korea, Republic of); Lee, Jang Su [Man-Machine Interface System team Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows.

  11. Research program plan: reactor vessels. Volume 1

    International Nuclear Information System (INIS)

    Vagins, M.; Taboada, A.

    1985-07-01

    The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

  12. US Department of Energy 1992--1993 Reactor Sharing Program

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1994-04-01

    The University of Florida Training Reactor serves as a host institution to support various educational institutions which are located primarily within the state of Florida. All users and uses were carefully screened to assure the usage was for educational institutions eligible for participation in the Reactor Sharing Program. Three tables are included that provide basic information about the 1992--1993 program and utilization of the reactor facilities by user institutions

  13. Reactor vessel using metal oxide ceramic membranes

    Science.gov (United States)

    Anderson, Marc A.; Zeltner, Walter A.

    1992-08-11

    A reaction vessel for use in photoelectrochemical reactions includes as its reactive surface a metal oxide porous ceramic membrane of a catalytic metal such as titanium. The reaction vessel includes a light source and a counter electrode. A provision for applying an electrical bias between the membrane and the counter electrode permits the Fermi levels of potential reaction to be favored so that certain reactions may be favored in the vessel. The electrical biasing is also useful for the cleaning of the catalytic membrane.

  14. SP-100 Program: space reactor system and subsystem investigations

    International Nuclear Information System (INIS)

    Harty, R.B.

    1983-01-01

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs

  15. Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

    International Nuclear Information System (INIS)

    Griffin, C.W.; Rasmussen, D.E.; Lloyd, M.H.

    1983-01-01

    The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities

  16. Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

    International Nuclear Information System (INIS)

    1975-06-01

    The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

  17. Development Program of the Advanced HANARO Reactor in Korea

    International Nuclear Information System (INIS)

    Yang, I.-S.; Ahn, J.-H.; Han, K.-I.; Parh, C.; Jun, B.-J.; Kim, Y.-J.

    2006-01-01

    The development program of an advanced HANARO (AHR) reactor started in Korea to keep abreast of the increasing future demand, from both home and abroad, for research activities. This paper provides a review of the status of research reactors in Korea, the operating experience of the HANARO, the design principles and preliminary features of an advanced HANARO reactor, and the specific strategy of an advanced HANARO reactor development program. The design principles were established in order to design a new multi-purpose research reactor that is safe, economically competitive and technically feasible. These include the adaptation of the HANARO design concept, its operating experience, a high ratio of flux to power, a high degree of safety, improved economic efficiency, improved operability and maintainability, increased space and expandability, and ALARA design optimization. The strategy of an advanced HANARO reactor development program considers items such as providing a digital advanced HANARO reactor in cyber space, a method for the improving the design quality and economy of research reactors by using Computer Integrated Engineering, and more effective advertising using diverse virtual reality. This development program will be useful for promoting the understanding of and interest in the operating HANARO as well as an advanced HANARO reactor under development in Korea. It will provide very useful information to a country that may need a research reactor in the near future for the promotion of public health, bio-technology, drug design, pharmacology, material processing, and the development of new materials. (author)

  18. Caramel, uranium oxide fuel plates for water cooled reactors

    International Nuclear Information System (INIS)

    Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

    1979-01-01

    The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

  19. Light Water Reactor Sustainability Program Integrated Program Plan

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  20. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-01

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

  1. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  2. Impact of Oxidative Stress in Fetal Programming

    OpenAIRE

    Thompson, Loren P.; Al-Hasan, Yazan

    2012-01-01

    Intrauterine stress induces increased risk of adult disease through fetal programming mechanisms. Oxidative stress can be generated by several conditions, such as, prenatal hypoxia, maternal under- and overnutrition, and excessive glucocorticoid exposure. The role of oxidant molecules as signaling factors in fetal programming via epigenetic mechanisms is discussed. By linking oxidative stress with dysregulation of specific target genes, we may be able to develop therapeutic strategies that pr...

  3. The development of fast simulation program for marine reactor parameters

    International Nuclear Information System (INIS)

    Chen Zhiyun; Hao Jianli; Chen Wenzhen

    2012-01-01

    Highlights: ► The simplified physical and mathematical models are proposed for a marine reactor system. ► A program is developed with Simulink module and Matlab file. ► The program developed has the merit of easy input preparation, output processing and fast running. ► The program can be used for the fast simulation of marine reactor parameters on the operating field. - Abstract: The fast simulation program for marine reactor parameters is developed based on the Simulink simulating software according to the characteristics of marine reactor with requirement of maneuverability and acute and fast response. The simplified core physical and thermal model, pressurizer model, steam generator model, control rod model, reactivity model and the corresponding Simulink modules are established. The whole program is developed by coupling all the Simulink modules. Two typical transient processes of marine reactor with fast load increase at low power level and load rejection at high power level are adopted to verify the program. The results are compared with those of Relap5/Mod3.2 with good consistency, and the program runs very fast. It is shown that the program is correct and suitable for the fast and accurate simulation of marine reactor parameters on the operating field, which is significant to the marine reactor safe operation.

  4. Comparison of advanced reactors program of different international vendors

    International Nuclear Information System (INIS)

    Agnihotri, N.K.

    2001-01-01

    The full text follows. Proposal for presenting a paper on Advanced Reactor Program Given below is the abstract for Track 6 session on Advanced Reactor at the ninth International Conference on Nuclear Engineering being held in Nice, France from April 8. through 12. 2001. This paper will provide an update on Advanced Reactor Program of different vendors in the United States, Japan, and Europe. Specifically the paper will look at the history of different Advanced Reactor Programs, international experience, aspect of economy due to standardization, and the highlights of technical specifications. The paper will also review aspects of Economy due to standardization, public acceptance, required construction time, and the experience of different vendors. The objective of the presentation is to underscore the highlights of the Reactor Program of different vendors in order to keep the attendees of the conference up-to-date. The presentation will be an impartial overview from an outsider's (not part of the Nuclear Steam Supply System's staff). (author)

  5. Non-linear programming method in optimization of fast reactors

    International Nuclear Information System (INIS)

    Pavelesku, M.; Dumitresku, Kh.; Adam, S.

    1975-01-01

    Application of the non-linear programming methods on optimization of nuclear materials distribution in fast reactor is discussed. The programming task composition is made on the basis of the reactor calculation dependent on the fuel distribution strategy. As an illustration of this method application the solution of simple example is given. Solution of the non-linear program is done on the basis of the numerical method SUMT. (I.T.)

  6. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Busby, Jeremy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barnard, Cathy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-01

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

  7. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-04-01

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

  8. Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

    International Nuclear Information System (INIS)

    Cowell, B.S.

    1997-06-01

    This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy

  9. Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.

    1991-11-01

    This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

  10. The national standards program for research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1977-01-01

    In 1970 a standards committee called ANS-15 was established by the American Nuclear Society (ANS) to prepare appropriate standards for research reactors. In addition, ANS acts as Secretariat for a national standards committee N17 which is responsible to the American National Standards Institute (ANSI) for the national consensus efforts for standards related to research reactors. To date ANS-15 has completed or is working on 14 standards covering all aspects of the operation of research reactors. Of the 11 research reactor standards submitted to the ANSI N17 Committee since its inception, six have been issued as National standards, and the remaining are still in the process of review. (author)

  11. Light Water Reactor Sustainability Program: Integrated Program Plan

    International Nuclear Information System (INIS)

    2016-02-01

    and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

  12. Light Water Reactor Sustainability Program: Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2016-02-15

    proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

  13. Participation in the US Department of Energy Reactor Sharing Program

    International Nuclear Information System (INIS)

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report

  14. Breeder reactor program in the USA

    International Nuclear Information System (INIS)

    Brewer, S.

    1978-01-01

    In the United States, commercial fuel reprocessing and demonstration test of plutonium breeder reactors were now postponed. LMFBR project and schedule of FFTF and afterwards await the results of INFCE. However, this is not discarding the development of LMFBRs. With the existing energy resources, the United States can have the opportunity to make careful and thorough study. Emphasis is placed on the research and development on new alternative types of fuel. FFTF going to be operated soon should provide effective means for the developments of FBR fuel and materials. High priority is to be retained for the test and development of sodium system hardwares. The nuclear proliferation problem is not related to heat transfer and secondary systems; it is associated with the selection of fuel and fuel cycle. The whole program is centered around LMFBR design and development. The target output will be 600 x 10 3 -- 700 x 10 3 MW. In the United States, now is the time to develop excellent products and to study the nuclear proliferation problem more carefully. (Mori, K.)

  15. Emergency medical assistance programs for nuclear power reactors

    International Nuclear Information System (INIS)

    Linnemann, R.E.; Mettler, F.A. Jr.

    1977-01-01

    This paper deals with a simple but practical medical support of geographically distributed nuclear reactors in isolated areas. A staff of experts at a centre devote their full attention to accident prevention and preparedness at reactor sites. They establish and maintain emergency medical programs at reactor sites and nearby support hospitals. The emphasis is on first aid and emergency treatment by medical attendants who are not and cannot be experts in radiation but do know how to treat patients. (author)

  16. DOE University Reactor Sharing Program. Renewal for 1994--1995

    International Nuclear Information System (INIS)

    Chappas, W.J.; Adams, V.G.

    1994-01-01

    The Department of Energy University Reactor Sharing Program at University of Maryland, College Park (UMCP) has, once again, stimulated a broad use of the reactor facilities by undergraduate and graduate students, visitors, and professionals. Participants are exposed to topics such as nuclear engineering, radiation safety, and nuclear reactor operations. This information is presented through various means including tours, slide presentations, experiments, and discussions. Student research using the MUTR is also encouraged. In addition, the Reactor Sharing Program here at the University of Maryland does not limit itself to the confines of the TRIGA reactor facility. Incorporated in the program are the Maryland University Neutron Activation Analysis Laboratory, the Maryland University Radiation Effects Laboratory, and the UMCP 2x4 Thermal Hydraulic Loop. These facilities enhance and give an added dimension to the tours and experiments

  17. Reliability of fast reactor mixed-oxide fuel during operational transients

    International Nuclear Information System (INIS)

    Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

    1991-07-01

    Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

  18. Light Water Reactor Sustainability Program: Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-05-01

    proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

  19. Reactor modeling and process analysis for partial oxidation of natural gas

    NARCIS (Netherlands)

    Albrecht, B.A.

    2004-01-01

    This thesis analyses a novel process of partial oxidation of natural gas and develops a numerical tool for the partial oxidation reactor modeling. The proposed process generates syngas in an integrated plant of a partial oxidation reactor, a syngas turbine and an air separation unit. This is called

  20. Advanced reactor development: The LMR integral fast reactor program at Argonne

    International Nuclear Information System (INIS)

    Till, C.E.

    1990-01-01

    Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

  1. The safety basis of the integral fast reactor program

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Seidel, B.R.

    1990-01-01

    The Integral Fast Reactor (IFR) and metallic fuel have emerged as the US Department of Energy reference reactor concept and fuel system for the development of an advanced liquid-metal reactor. This article addresses the basic elements of the IFR reactor concept and focuses on the safety advances achieved by the IFR Program in the areas of (1) fuel performance, (2) superior local faults tolerance, (3) transient fuel performance, (4) fuel-failure mechanisms, (5) performance in anticipated transients without scram, (6) core-melt mitigation, and (7) actinide recycle

  2. Method and program for complex calculation of heterogeneous reactor

    International Nuclear Information System (INIS)

    Kalashnikov, A.G.; Glebov, A.P.; Elovskaya, L.F.; Kuznetsova, L.I.

    1988-01-01

    An algorithm and the GITA program for complex one-dimensional calculation of a heterogeneous reactor which permits to conduct calculations for the reactor and its cell simultaneously using the same algorithm are described. Multigroup macrocross sections for reactor zones in the thermal energy range are determined according to the technique for calculating a cell with complicate structure and then the continuous multi group calculation of the reactor in the thermal energy range and in the range of neutron thermalization is made. The kinetic equation is solved using the Pi- and DSn- approximations [fr

  3. The chemical energy unit partial oxidation reactor operation simulation modeling

    Science.gov (United States)

    Mrakin, A. N.; Selivanov, A. A.; Batrakov, P. A.; Sotnikov, D. G.

    2018-01-01

    The chemical energy unit scheme for synthesis gas, electric and heat energy production which is possible to be used both for the chemical industry on-site facilities and under field conditions is represented in the paper. The partial oxidation reactor gasification process mathematical model is described and reaction products composition and temperature determining algorithm flow diagram is shown. The developed software product verification showed good convergence of the experimental values and calculations according to the other programmes: the temperature determining relative discrepancy amounted from 4 to 5 %, while the absolute composition discrepancy ranged from 1 to 3%. The synthesis gas composition was found out practically not to depend on the supplied into the partial oxidation reactor (POR) water vapour enthalpy and compressor air pressure increase ratio. Moreover, air consumption coefficient α increase from 0.7 to 0.9 was found out to decrease synthesis gas target components (carbon and hydrogen oxides) specific yield by nearly 2 times and synthesis gas target components required ratio was revealed to be seen in the water vapour specific consumption area (from 5 to 6 kg/kg of fuel).

  4. Results and recommendations from the reactor chemistry and corrosion tasks of the reactor materials program

    International Nuclear Information System (INIS)

    Baumann, E.W.; Ondrejcin, R.S.

    1990-11-01

    Within the general context of extended service life, the Reactor Materials Program was initiated in 1984. This comprehensive program addressed material performance in SRS reactor tanks and the primary coolant or Process Water System (PWS) piping. Three of the eleven tasks concerned moderator quality and corrosion mitigation. Definition and control of the stainless steel aqueous environment is a key factor in corrosion mitigation. The Reactor Materials Program systematically investigated the SRS environment and its effect on crack initiation and propagation in stainless steel, with the objective of improving this environment. The purpose of this report is to summarize the contributions of Tasks 6, 7 and 10 of the Reactor Materials Program to the understanding and control of moderator quality and its relationship to mitigation of stress corrosion cracking

  5. A Prediction Study of Aluminum Alloy Oxidation of the Fuel Cladding in Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tahk, Y. W.; Oh, J. Y.; Lee, B. H.; Seo, C. G.; Chae, H. T.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    U{sub 3}Si{sub 2}-Al dispersion fuel with Al cladding will be used for Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding experiences the oxidation layer growth on the surface during the reactor operation. The formation of oxides on the cladding affects fuel performance by increasing fuel temperature. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, a fresh fuel is discharged after 900 effective full power days (EFPD) with 18 cycles of 50 days loading. For the proper prediction of the aluminum oxide thickness of fuel cladding during the long residence time, a reliable model is needed. In this work, several oxide thickness prediction models are compared with the measured data from in-pile test by RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model are performed for JRTR fuel

  6. Oxide production program monthly report - December 2014

    Energy Technology Data Exchange (ETDEWEB)

    Kelley, Evelyn A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Whitworth, Julia [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lloyd, Jane Alexandria [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hampton, David Earl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Benavidez, Amelia A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-15

    A summary of the major activities, accomplishments, milestones, financial summary, project performance and issues facing the ARIES Oxide Production Program for the month of December 2014 is presented in this Executive Summary.

  7. Decentralization of operating reactor licensing reviews: NRR Pilot Program

    International Nuclear Information System (INIS)

    Hannon, J.N.

    1984-07-01

    This report, which has incorporated comments received from the Commission and ACRS, describes the program for decentralization of selected operating reactor licensing technical review activities. The 2-year pilot program will be reviewed to verify that safety is enhanced as anticipated by the incorporation of prescribed management techniques and application of resources. If the program fails to operate as designed, it will be terminated

  8. Computerised programming of the Dragon reactor fuel handling operations

    International Nuclear Information System (INIS)

    Butcher, P.

    1976-11-01

    Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

  9. The UMR reactor outreach program for expanded educational utilization

    International Nuclear Information System (INIS)

    Freeman, D.; Bolon, A.

    1992-01-01

    In recent years, the University of Missouri-Rolla Reactor (UMRR) facility has been under intense financial scrutiny by the university administration; primarily due to ever-tightening budgets and declines in nuclear engineering (NE) enrollment. In response to criticisms of low utilization, the reactor staff has developed and implemented a dynamic outreach program designed to significantly increase the educational role of the facility on campus. The outreach program is based on the principle that the potential to provide service to the UMR community is far in excess of the present level of service. The program is designed to identify and inform potential users of how their courses or programs can be augmented through use of the reactor facility. The net effect of the outreach program is greater campus communication and awareness of the unique capabilities as applied to each discipline. A natural product of the outreach program should be increased research

  10. A model for oxidizing species concentrations in boiling water reactors

    International Nuclear Information System (INIS)

    Sun, B.; Chexal, B.; Pathania, R.; Chun, J.; Ballinger, R.; Abdollahian, D.

    1993-01-01

    To evaluate and control the intergranular stress corrosion cracking of boiling water reactor (BWR) vessel internal components requires knowledge of the concentration of oxidizing species that affects the electrochemical potentials in various regions of a BWR. In a BWR flow circuit, as water flows through the radiation field, the radiolysis process and chemical reactions lead to the production of species such as oxygen, hydrogen, and hydrogen peroxide. Since chemistry measurements are difficult inside BWRs, analytical tools have been developed by Ruiz and Lin, Ibe and Uchida and Chun and Ballinger for estimating the concentration of species that provide the necessary input for water chemistry control and material protection

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

    International Nuclear Information System (INIS)

    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

  12. A program for dynamic noise investigations of reactor systems

    International Nuclear Information System (INIS)

    Antonov, N.A.; Yaneva, N.B.

    1980-01-01

    A stochastic process analysis in nuclear reactors is used for the state diagnosis and dynamic characteristic investigation of the reactor system. A program DENSITY adapted and tested on an IBM 360 ES type computer is developed. The program is adjusted for fast processing of long series exploiting a relatively small memory. The testing procedure is discussed and the method of the periodic sequences corresponding to characteristic reactivity perturbations of the reactor systems is considered. The program is written for calculating the auto-power spectral density and the cross-power spectral density, as well as the coherence function of stationary statistical time series using the advantages of the fast Fourier transformation. In particular, it is shown that the multi-frequency binary sequences are very useful with respect to the signal-to-noise ratio and the frequency distribution in view of the frequency reactor test

  13. OSU Reactor Sharing Program FY 1995 annual report

    International Nuclear Information System (INIS)

    Higginbotham, J.F.

    1996-10-01

    This is the annual report of the activities supported under the Oregon State University Reactor Sharing Program, award number DE-FG06-NE38137. The beginning date for the award was September, 30, 1995 and the end date was September 29, 1996. Work conducted under this award is internally administered at the Radiation Center through a project tasking system. This allows for excellent quality control for the work which is performed from the point of initial contact, through the reactor application, project report generation and financial accounting. For the current fiscal year, FY95, the total cost of the reactor sharing program, including Radiation Center contributions, was $66,323.20 of which $40,000.00 was supplied by the DOE Reactor Sharing Program. The details of individual project costs is given in Table 1. The work performed for the individual projects are described in the brief work descriptions given in Table 2

  14. Impact of VOC Composition and Reactor Conditions on the Aging of Biomass Cookstove Emissions in an Oxidation Flow Reactor

    Science.gov (United States)

    Oxidation flow reactor (OFR) experiments in our lab have explored secondary organic aerosol (SOA) production during photochemical aging of emissions from cookstoves used by billions in developing countries. Previous experiments, conducted with red oak fuel under conditions of hig...

  15. Fusion reactor design and technology program in China

    International Nuclear Information System (INIS)

    Huang, J.H.

    1994-01-01

    A fusion-fission hybrid reactor program was launched in 1987. The purpose of development of the hybrid reactor is twofold: to solve the problem of nuclear fuel supply for an expected large-scale development of fission reactor plants, and to maintain the momentum of fusion research. The program is described and the activities and progress of the program are presented. Two conceptual designs of an engineering test reactor with tokamak configuration were developed at the Southwestern Institute of Physics and the Institute of Plasma Physics. The results are a tokamak engineering test breeder (TETB) series design and a fusion-fission hybrid reactor design (SSEHR), characterized by a liquid-Li self-cooled blanket and an He-cooled solid tritium breeder blanket respectively. In parallel with the design studies, relevant technological experiments on a small or medium scale have been supported by this program. These include LHCD, ICRH and pellet injection in the area of plasma engineering; neutronics integral experiments with U, Pu, Fe and Be; various irradiation tests of austenitic and ferritic steels, magnetohydrodynamic (MHD) pressure drop experiments using a liquid metal loop; research into permeation barriers for tritium and hydrogen isotopes; solid tritium breeder tests using an in-situ loop in a fission reactor. All these experiments have proceeded successfully. The second step of this program is now starting. It seems reasonable that most of the research carried out in the first step will continue. ((orig.))

  16. Effect of reactor heat transfer limitations on CO preferential oxidation

    Science.gov (United States)

    Ouyang, X.; Besser, R. S.

    Our recent studies of CO preferential oxidation (PrOx) identified systematic differences between the characteristic curves of CO conversion for a microchannel reactor with thin-film wall catalyst and conventional mini packed-bed lab reactors (m-PBR's). Strong evidence has suggested that the reverse water-gas-shift (r-WGS) side reaction activated by temperature gradients in m-PBR's is the source of these differences. In the present work, a quasi-3D tubular non-isothermal reactor model based on the finite difference method was constructed to quantitatively study the effect of heat transport resistance on PrOx reaction behavior. First, the kinetic expressions for the three principal reactions involved were formed based on the combination of experimental data and literature reports and their parameters were evaluated with a non-linear regression method. Based on the resulting kinetic model and an energy balance derived for PrOx, the finite difference method was then adopted for the quasi-3D model. This model was then used to simulate both the microreactor and m-PBR's and to gain insights into their different conversion behavior. Simulation showed that the temperature gradients in m-PBR's favor the reverse water-gas-shift (r-WGS) reaction, thus causing a much narrower range of permissible operating temperature compared to the microreactor. Accordingly, the extremely efficient heat removal of the microchannel/thin-film catalyst system eliminates temperature gradients and efficiently prevents the onset of the r-WGS reaction.

  17. Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-02-01

    An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

  18. Low temperature oxidation of hydrocarbons using an electrochemical reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide

    conversion was a complex function of multiple variables: the microstructure of the backbone, the polarization resistance of the electrodes, both at OCV and under polarization, the electrical and morphological properties of the infiltrated material and the specific reaction conditions like the propene......This study investigated the use of a ceramic porous electrochemical reactor for the deep oxidation of propene. Two electrode composites, La0.85Sr0.15MnO3±d/Ce0.9Gd0.1O1.95 (LSM/CGO) and La0.85Sr0.15FeMnO3/Ce0.9Gd0.1O1.95 (LSF/CGO), were produced in a 5 single cells stacked configuration and used...... prolonged polarization was able to partially counteract the instability of the infiltrated Ce0.9Gd0.1O1.95. This project demonstrated the possibility to enhance the oxidation of propene by polarization in a porous ceramic reactor. The infiltration of different active materials helped to increase...

  19. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    International Nuclear Information System (INIS)

    Abe, K.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    1998-01-01

    A Japan-USA program of irradiation experiments for fusion research, ''JUPITER'', has been established as a 6 year program from 1995 to 2000. The goal is to study ''the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment''. This is phase-three of the collaborative program, which follows RTNS-II program (phase-1: 1982-1986) and FFTF/MOTA program (phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA program. JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. (orig.)

  20. Materials surveillance program for C-E NSSS reactor vessels

    International Nuclear Information System (INIS)

    Koziol, J.J.

    1977-01-01

    Irradiation surveillance programs for light water NSSS reactor vessels provide the means by which the utility can assess the extent of neutron-induced changes in the reactor vessel materials. These programs are conducted to verify, by direct measurement, the conservatism in the predicted radiation-induced changes and hence the operational parameters (i.e., heat-up, cooldown, and pressurization rates). In addition, such programs provide assurance that the scheduled adjustments in the operational parameters are made with ample margin for safe operation of the plant. During the past 3 years, several documents have been promulgated establishing the criteria for determining both the initial properties of the reactor vessel materials as well as measurement of changes in these initial properties as a result of irradiation. These documents, ASTM E-185-73, ''Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,'' and Appendix H to 10 CFR 50, ''Reactor Vessel Material Surveillance Program Requirements,'' are complementary to each other. They are the result of a change in the basic philosophy regarding the design and analysis of reactor vessels. In effect, the empirical ''transition temperature approach,'' which was used for design, was replaced by the ''analytical fracture mechanics approach.'' The implementation of this technique was described in Welding Research Council Bulletin 1975 and Appendix G to ASME Code Section III. Further definition of requirements appears in Appendix G to 10 CFR 50 published in July 1973. It is the intent of this paper to describe (1) a typical materials surveillance program for the reactor vessel of a Combustion Engineering NSSS, and (2) how the results of such programs, as well as experimental programs provide feed-back for improvement of materials to enhance their radiation resistance and thereby further improve the safety and reliability of future plants. (author)

  1. Supercritical Water Oxidation Program (SCWOP)

    International Nuclear Information System (INIS)

    1994-02-01

    Purpose of SCWOP is to develop and demonstrate supercritical water oxidation as a viable technology for treating DOE hazardous and mixed wastes and to coordinate SCWO research, development, demonstration, testing, and evaluation activities. The process involves bringing together organic waste, water, and an oxidant (air, O 2 , etc.) to temperatures and pressures above water's critical point (374 C, 22.1 MPa); organic destruction is >99.99% efficient, and the resulting effluents (mostly water, CO 2 ) are relatively benign. Pilot-scale (300--500 gallons/day) SCWO units are to be constructed and demonstrated. Two phases will be conducted: hazardous waste pilot plant demonstration and mixed waste pilot demonstration. Contacts for further information and for getting involved are given

  2. Safety program considerations for space nuclear reactor systems

    International Nuclear Information System (INIS)

    Cropp, L.O.

    1984-08-01

    This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given

  3. Westinghouse Small Modular Reactor (SMR) Programe

    International Nuclear Information System (INIS)

    Shulyak, Nick

    2014-01-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) in which all primarycomponents associated with the nuclear steam supply system, including the steam generator and the pressurizer, are housed within the reactor vessel. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. This paper describes the design and functionality of the Westinghouse SMR, the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design drivers include safety, economics, reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 reactor, and provides mitigation of all design basis accidents without the need for offsite AC electrical power for a period of seven days. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reactors (LWR) that utilized the economies of scale to reach economic competiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is implemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology. The integral Westinghouse SMR design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high

  4. Determination of melting point of mixed-oxide fuel irradiated in a fast breeder reactor

    International Nuclear Information System (INIS)

    Tachibana, Toshimichi

    1985-01-01

    The melting point of fuel is important to set its in-reactor maximum temperature in fuel design. The fuel melting point measuring methods are broadly the filament method and the capsule sealing method. The only instance of measuring the melting point of irradiated mixed oxide (U, Pu)O 2 fuel by the filament method is by GE in the United States. The capsule sealing method, while the excellent means, is difficult in weld sealing the irradiated fuel in a capsule within the cell. In the fast reactor development program, the remotely operated melting point measuring apparatus in capsule sealing the mixed (U, Pu)O 2 fuel irradiated in the experimental FBR Joyo was set in the cell and the melting point was measured, for the first time in the world. (Mori, K.)

  5. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Yoshida, A.; Yogo, S.

    2001-01-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  6. FISS: a computer program for reactor systems studies

    International Nuclear Information System (INIS)

    Tamm, H.; Sherman, G.R.; Wright, J.H.; Nieman, R.E.

    1979-08-01

    ΣFISSΣ is a computer code for use in investigating alternative fuel cycle strategies for Canadian and world nuclear programs. The code performs a system simulation accounting for dynamic effects of growing nuclear systems. Facilities in the model include storage for irradiated fuel, mines, plants for enrichment, fuel fabrication, fuel reprocessing and heavy water, and reactors. FISS is particularly useful for comparing various reactor strategies and studying sensitivities of resource consumption, capital investment and energy costs with changes in fuel cycle parameters, reactor parameters and financial variables. (author)

  7. Reactor physical experimental program EROS in the frame of the molten salt applying reactor concepts development

    International Nuclear Information System (INIS)

    Hron, Miloslav; Kyncl, Jan; Mikisek, Miroslav

    2009-01-01

    After the relatively broad program of experimental activities, which have been involved in the complex R and D program for the Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept development in the Czech Republic, there has been a next stage (namely large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0 called EROS project) started, which will be focused to the experimental verification of the rector physical or neutronic properties of other types of reactor concepts applying molten salts in the role of liquid fuel and/or coolant. This tendency is based on the recently accepted decision of the MSR SSC of GIF to consider for further period of its activity two baseline concepts- fast neutron molten salt reactor non-moderated (FMSR-NM) as a long-term alternative to solid fuelled fast neutron reactors and simultaneously, advanced high temperature reactor (AHTR) with pebble bed type solid fuel cooled by liquid salts. There will be a brief description of the prepared and performed experimental programs in these directions (as well as the preliminary results obtained so far) introduced in the paper. (author)

  8. Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

    International Nuclear Information System (INIS)

    Hoover, M.D.; Newton, G.J.

    1997-01-01

    Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

  9. Research and development program in reactor safety for NUCLEBRAS

    International Nuclear Information System (INIS)

    Pinheiro, R.B.; Resende Lobo, A.A. de; Horta, J.A.L.; Avelar Esteves, F. de; Lepecki, W.P.S.; Mohr, K.; Selvatici, E.

    1984-01-01

    With technical assistance from the IAEA, it was established recently an analytical and experimental Research and Development Program for NUCLEBRAS in the area of reactor safety. The main objectives of this program is to make possible, with low investments, the active participation of NUCLEBRAS in international PWR safety research. The analytical and experimental activities of the program are described with some detail, and the main results achieved up to now are presented. (Author) [pt

  10. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

    2004-01-01

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  11. Improving nuclear safety at international research reactors: The Integrated Research Reactor Safety Enhancement Program (IRRSEP)

    International Nuclear Information System (INIS)

    Huizenga, David; Newton, Douglas; Connery, Joyce

    2002-01-01

    Nuclear energy continues to play a major role in the world's energy economy. Research and test reactors are an important component of a nation's nuclear power infrastructure as they provide training, experiments and operating experience vital to developing and sustaining the industry. Indeed, nations with aspirations for nuclear power development usually begin their programs with a research reactor program. Research reactors also are vital to international science and technology development. It is important to keep them safe from both accident and sabotage, not only because of our obligation to prevent human and environmental consequence but also to prevent corresponding damage to science and industry. For example, an incident at a research reactor could cause a political and public backlash that would do irreparable harm to national nuclear programs. Following the accidents at Three Mile Island and Chernobyl, considerable efforts and resources were committed to improving the safety posture of the world's nuclear power plants. Unsafe operation of research reactors will have an amplifying effect throughout a country or region's entire nuclear programs due to political, economic and nuclear infrastructure consequences. (author)

  12. Development of a simultaneous partial nitrification and anaerobic ammonia oxidation process in a single reactor.

    Science.gov (United States)

    Cho, Sunja; Fujii, Naoki; Lee, Taeho; Okabe, Satoshi

    2011-01-01

    Up-flow oxygen-controlled biofilm reactors equipped with a non-woven fabric support were used as a single reactor system for autotrophic nitrogen removal based on a combined partial nitrification and anaerobic ammonium oxidation (anammox) reaction. The up-flow biofilm reactors were initiated as either a partial nitrifying reactor or an anammox reactor, respectively, and simultaneous partial nitrification and anammox was established by careful control of the aeration rate. The combined partial nitrification and anammox reaction was successfully developed in both biofilm reactors without additional biomass inoculation. The reactor initiated as the anammox reactor gave a slightly higher and more stable mean nitrogen removal rate of 0.35 (±0.19) kg-N m(-3) d(-1) than the reactor initiated as the partial nitrifying reactor (0.23 (±0.16) kg-N m(-3) d(-1)). FISH analysis revealed that the biofilm in the reactor started as the anammox reactor were composed of anammox bacteria located in inner anoxic layers that were surrounded by surface aerobic AOB layers, whereas AOB and anammox bacteria were mixed without a distinguishable niche in the biofilm in the reactor started as the partial nitrifying reactor. However, it was difficult to efficiently maintain the stable partial nitrification owing to inefficient aeration in the reactor, which is a key to development of the combined partial nitrification and anammox reaction in a single biofilm reactor. Copyright © 2010 Elsevier Ltd. All rights reserved.

  13. Enhanced performance of solid oxide electrolysis cells by integration with a partial oxidation reactor: Energy and exergy analyses

    International Nuclear Information System (INIS)

    Visitdumrongkul, Nuttawut; Tippawan, Phanicha; Authayanun, Suthida; Assabumrungrat, Suttichai; Arpornwichanop, Amornchai

    2016-01-01

    Highlights: • Process design of solid oxide electrolyzer integrated with a partial oxidation reactor is studied. • Effect of key operating parameters of partial oxidation reactor on the electrolyzer performance is presented. • Exergy analysis of the electrolyzer process is performed. • Partial oxidation reactor can enhance the solid oxide electrolyzer performance. • Partial oxidation reactor in the process is the highest exergy destruction unit. - Abstract: Hydrogen production without carbon dioxide emission has received a large amount of attention recently. A solid oxide electrolysis cell (SOEC) can produce pure hydrogen and oxygen via a steam electrolysis reaction that does not emit greenhouse gases. Due to the high operating temperature of SOEC, an external heat source is required for operation, which also helps to improve SOEC performance and reduce operating electricity. The non-catalytic partial oxidation reaction (POX), which is a highly exothermic reaction, can be used as an external heat source and can be integrated with SOEC. Therefore, the aim of this work is to study the effect of operating parameters of non-catalytic POX (i.e., the oxygen to carbon ratio, operating temperature and pressure) on SOEC performance, including exergy analysis of the process. The study indicates that non-catalytic partial oxidation can enhance the hydrogen production rate and efficiency of the system. In terms of exergy analysis, the non-catalytic partial oxidation reactor is demonstrated to be the highest exergy destruction unit due to irreversible chemical reactions taking place, whereas SOEC is a low exergy destruction unit. This result indicates that the partial oxidation reactor should be improved and optimally designed to obtain a high energy and exergy system efficiency.

  14. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  15. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    1994-01-01

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

  16. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  17. In-core fuel management programs for nuclear power reactors

    International Nuclear Information System (INIS)

    1984-10-01

    In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

  18. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  19. Pacific Northwest Laboratory Monthly Activities Report for August 1966 AEC Division of Reactor Development and Technology Programs

    Energy Technology Data Exchange (ETDEWEB)

    SL Fawcett

    1966-08-01

    This report has the following sections: Summary; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; and Nuclear Safety.

  20. A flow reactor for the flow supercritical water oxidation of wastes to mitigate the reactor corrosion problem

    International Nuclear Information System (INIS)

    Chitanvis, S.M.

    1994-01-01

    We have designed a flow tube reactor for supercritical water oxidation of wastes that confines the oxidation reaction to the vicinity of the axis of the tube. This prevents high temperatures and reactants as well as reaction products from coming in intimate contact with reactor walls. This implies a lessening of corrosion of the walls of the reactor. We display numerical simulations for a vertical reactor with conservative design parameters that illustrate our concept. We performed our calculations for the destruction of sodium nitrate by ammonium hydroxide In the presence of supercritical water, where the production of sodium hydroxide causes corrosion. We have compared these results with that for a horizontal set-up where the sodium hydroxide created during the reaction ends up on the floor of the tube, implying a higher probability of corrosion

  1. Biological oxidation of dissolved methane in effluents from anaerobic reactors using a down-flow hanging sponge reactor.

    Science.gov (United States)

    Hatamoto, Masashi; Yamamoto, Hiroki; Kindaichi, Tomonori; Ozaki, Noriatsu; Ohashi, Akiyoshi

    2010-03-01

    Anaerobic wastewater treatment plants discharge dissolved methane, which is usually not recovered. To prevent emission of methane, which is a greenhouse gas, we utilized an encapsulated down-flow hanging sponge reactor as a post-treatment to biologically oxidize dissolved methane. Within 3 weeks after reactor start-up, methane removal efficiency of up to 95% was achieved with a methane removal rate of 0.8 kg COD m(-3) day(-1) at an HRT of 2 h. After increasing the methane-loading rate, the maximum methane removal rate reached 2.2 kg COD m(-3) day(-1) at an HRT of 0.5 h. On the other hand, only about 10% of influent ammonium was oxidized to nitrate during the first period, but as airflow was increased to 2.5 L day(-1), nitrification efficiency increased to approximately 70%. However, the ammonia oxidation rate then decreased with an increase in the methane-loading rate. These results indicate that methane oxidation occurred preferentially over ammonium oxidation in the reactor. Cloning of the 16S rRNA and pmoA genes as well as phylogenetic and T-RFLP analyses revealed that type I methanotrophs were the dominant methane oxidizers, whereas type II methanotrophs were detected only in minor portion of the reactor. Copyright 2009 Elsevier Ltd. All rights reserved.

  2. Inspection program for U.S. research reactors

    International Nuclear Information System (INIS)

    Isaac, Patrick J.

    2010-01-01

    This paper presents an established program for inspection of nuclear research reactors to ensure that systems and techniques are in accordance with regulatory requirements and to provide protection for the health and safety of the public. The inspection program, implemented from the time a facility gets licensed, remains in effect through operations, shutdown, decommissioning, and until the license is terminated. The program establishes inspection methodology for operating, safeguards, and decommissioning activities. Using a performance- based approach, inspectors focus their attention on activities important to safety. Inspection procedures allow the inspectors to assess facility safety and compliance to applicable requirements. A well designed inspection program is an integral part of the mechanism to ensure that the level of performance in the strategic areas of reactor safety, radiation safety, and safeguards is acceptable and provides adequate protection of public health and safety. (author)

  3. US Advanced Light Water Reactor Program; overall objective

    International Nuclear Information System (INIS)

    Klug, N.

    1989-01-01

    The overall objective of the US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) program is to perform coordinated programs of the nuclear industry and DOE to insure the availability of licensed, improved, and simplified light water reactor standard plant designs that may be ordered in the 1990's to help meet the US electrical power demand. The discussion includes plans to meet program objectives and the design certification program. DOE is currently supporting the development of conceptual designs, configurations, arrangements, construction methods/plans, and proof test key design features for the General Electric ASBWR and the Westinghouse AP600. Key features of each are summarized. Principal milestones related to licensing of large standard plants, simplified mid-size plant development, and plant lifetime improvement are noted

  4. Fusion Reactor Safety Research Program annual report, FY-79

    International Nuclear Information System (INIS)

    Crocker, J.G.; Cohen, S.

    1980-08-01

    The objective of the program is the development, coordination, and execution of activities related to magnetic fusion devices and reactors that will: (a) identify and evaluate potential hazards, (b) assess and disclose potential environmental impacts, and (c) develop design standards and criteria that eliminate, mitigate, or reduce those hazards and impacts. The program will provide a sound basis for licensing fusion reactors. Included in this report are portions of four reports from two outside contractors, discussions of the several areas in which EG and G Idaho is conducting research activities, a discussion of proposed program plan development, mention of special tasks, a review of fusion technology program coordination by EG and G with other laboratories, and a brief view of proposed FY-80 activities

  5. Integral Fast Reactor Program annual progress report, FY 1991

    International Nuclear Information System (INIS)

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  6. Integral Fast Reactor Program. Annual progress report, FY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  7. Integral Fast Reactor Program annual progress report, FY 1994

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, J.J.

    1994-12-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R ampersand D

  8. Integral Fast Reactor Program. Annual progress report, FY 1993

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D.

  9. Integral Fast Reactor Program. Annual progress report, FY 1993

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D

  10. Japanese program of materials research for fusion reactors

    International Nuclear Information System (INIS)

    Hasiguti, R.R.

    1982-01-01

    The Japanese program of materials research for fusion reactors is described based on the report to the Nuclear Fusion Council, the project research program of the Ministry of Education, Science and Culture, and other official documents. The alloy development for the first wall and its radiation damage are the main topics discussed in this paper. Materials viewpoints for the Japanese Tokamak facilities and the problems of irradiation facilities are also discussed. (orig.)

  11. Users guide to the computer program FURST (FUture Reactor STrategies)

    International Nuclear Information System (INIS)

    Hatton, H.

    1981-01-01

    A program has been written to calculate the future resource requirements for the nuclear portion of an electricity generating system. Starting from a given total energy demand projection the program calculates the required growth of the electrical generating system, the total nuclear system, and the portion provided by reactors using an advanced fuel cycle by successive application of the Fisher/Pry penetration formula. Several options are available. These include the ability to (1) change the growth rates of any part of the system; (2) change the characteristics of the reactors; (3) include the effects of decommissioning reactors at the end of their design lifetimes; (4) vary the date of introduction of advanced reactors; and (5) limit the amount of natural uranium available annually. The output gives the history of the growth of the nuclear system and the uranium mining and fuel reprocessing requirements. The output can be obtained either as tables of numbers or graphs with crossplots to compare reactor systems or total energy scenarios. (author)

  12. Assessment of specialized educational programs for licensed nuclear reactor operators

    International Nuclear Information System (INIS)

    Melber, B.D.; Saari, L.M.; White, A.S.; Geisendorfer, C.L.; Huenefeld, J.C.

    1986-02-01

    This report assesses the job-relatedness of specialized educational programs for licensed nuclear reactor operators. The approach used involved systematically comparing the curriculum of specialized educational programs for college credit, to academic knowledge identified as necessary for carrying out the jobs of licenses reactor operators. A sample of eight programs, including A.S. degree, B.S. degree, and coursework programs were studied. Subject matter experts in the field of nuclear operations curriculum and training determined the extent to which individual program curricula covered the identified job-related academic knowledge. The major conclusions of the report are: There is a great deal of variation among individual programs, ranging from coverage of 15% to 65% of the job-related academic knowledge. Four schools cover at least half, and four schools cover less than one-third of this knowledge content; There is no systematic difference in the job-relatedness of the different types of specialized educational programs, A.S. degree, B.S. degree, and coursework; and Traditional B.S. degree programs in nuclear engineering cover as much job-related knowledge (about one-half of this knowledge content) as most of the specialized educational programs

  13. Nuclear Safety Research Reactor (NSRR) program in JAERI

    International Nuclear Information System (INIS)

    Ishikawa, M.; Hoshi, T.; Ohnishi, N.; Fujishiro, T.; Inabe, T.

    1974-01-01

    An experimental research program, named Nuclear Safety Research Reactor (NSRR) Program, has been progressing in Japan Atomic Energy Research Institute (JAERI) using a modified TRIGA-ACPR. This paper is prepared to describe the outline of the NSRR program. The purpose of the NSRR program is to examine the behaviors of fuel rods under various accidental conditions of power reactors so as to establish realistic safety criteria and to develop analytical models for prediction of fuel failures. We expect to contribute finally to the improvement of reactor design and fuel fabrication techniques based on these experimental results. The NSRR experiments will be performed in the large central experimental tube, which is one of the most excellent features of this reactor, using specially designed capsules or loops which can accommodate up to 49 BWR type test fuels. Many types of test fuels in various conditions will be examined by the NSRR program, such as BWR, PWR and FBR type fuels from the beginning of life to the end of life with and without simulated reactor internal structures. The experiments will be continued for more than 10 years divided into three phases. The first phase of the program will be devoted to the experiments pertaining to reactivity initiated accidents (RIA). In these experiments we will make use of the excellent pulsing capability of ACPR, which is expected to generate 100 MW-sec prompt energy release with 1.3 msec of minimum reactor period by 4.7 dollar reactivity insertion and to yield more than 280 cal/g-UO 2 heat deposit even in an approximately 10% enriched BWR type test fuel. (280 cal/g-UO 2 is believed enough heat deposit to cause fuel failure.) In general, heat flow behaviors from fuel meat to clad and from clad to coolant are very complex phenomena, but they are the key point in analyzing transient response of fuels. In the sudden heat transient conditions brought by pulsing, however, it will be possible to examine each phenomenon separately

  14. Nuclear Safety Research Reactor (NSRR) program in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, M; Hoshi, T; Ohnishi, N; Fujishiro, T; Inabe, T [Japan Atomic Energy Research Institute (Japan)

    1974-07-01

    An experimental research program, named Nuclear Safety Research Reactor (NSRR) Program, has been progressing in Japan Atomic Energy Research Institute (JAERI) using a modified TRIGA-ACPR. This paper is prepared to describe the outline of the NSRR program. The purpose of the NSRR program is to examine the behaviors of fuel rods under various accidental conditions of power reactors so as to establish realistic safety criteria and to develop analytical models for prediction of fuel failures. We expect to contribute finally to the improvement of reactor design and fuel fabrication techniques based on these experimental results. The NSRR experiments will be performed in the large central experimental tube, which is one of the most excellent features of this reactor, using specially designed capsules or loops which can accommodate up to 49 BWR type test fuels. Many types of test fuels in various conditions will be examined by the NSRR program, such as BWR, PWR and FBR type fuels from the beginning of life to the end of life with and without simulated reactor internal structures. The experiments will be continued for more than 10 years divided into three phases. The first phase of the program will be devoted to the experiments pertaining to reactivity initiated accidents (RIA). In these experiments we will make use of the excellent pulsing capability of ACPR, which is expected to generate 100 MW-sec prompt energy release with 1.3 msec of minimum reactor period by 4.7 dollar reactivity insertion and to yield more than 280 cal/g-UO{sub 2} heat deposit even in an approximately 10% enriched BWR type test fuel. (280 cal/g-UO{sub 2} is believed enough heat deposit to cause fuel failure.) In general, heat flow behaviors from fuel meat to clad and from clad to coolant are very complex phenomena, but they are the key point in analyzing transient response of fuels. In the sudden heat transient conditions brought by pulsing, however, it will be possible to examine each phenomenon

  15. Electrochemical oxidation of phenol in a parallel plate reactor using ruthenium mixed metal oxide electrode

    Energy Technology Data Exchange (ETDEWEB)

    Yavuz, Yusuf [Anadolu Universitesi, Cevre Sor. Uyg. ve Aras. Merkezi, Eskisehir (Turkey); Koparal, A. Savas [Anadolu Universitesi, Cevre Sor. Uyg. ve Aras. Merkezi, Eskisehir (Turkey)]. E-mail: askopara@anadolu.edu.tr

    2006-08-21

    In this study, electrochemical oxidation of phenol was carried out in a parallel plate reactor using ruthenium mixed metal oxide electrode. The effects of initial pH, temperature, supporting electrolyte concentration, current density, flow rate and initial phenol concentration on the removal efficiency were investigated. Model wastewater prepared with distilled water and phenol, was recirculated to the electrochemical reactor by a peristaltic pump. Sodium sulfate was used as supporting electrolyte. The Microtox'' (registered) bioassay was also used to measure the toxicity of the model wastewater during the study. As a result of the study, removal efficiency of 99.7% and 88.9% were achieved for the initial phenol concentration of 200 mg/L and chemical oxygen demand (COD) of 480 mg/L, respectively. In the same study, specific energy consumption of 1.88 kWh/g phenol removed and, mass transfer coefficient of 8.62 x 10{sup -6} m/s were reached at the current density of 15 mA/cm{sup 2}. Electrochemical oxygen demand (EOD), which can be defined as the amount of electrochemically formed oxygen used for the oxidation of organic pollutants, was 2.13 g O{sub 2}/g phenol. Electrochemical oxidation of petroleum refinery wastewater was also studied at the optimum experimental conditions obtained. Phenol removal of 94.5% and COD removal of 70.1% were reached at the current density of 20 mA/cm{sup 2} for the petroleum refinery wastewater.

  16. Status of Away From Reactor spent fuel storage program

    International Nuclear Information System (INIS)

    King, F.D.

    1979-07-01

    The Away From Reactor (AFR) Spent Fuel Program that the US Department of Energy established in 1977 is intended to preclude the shutting down of commercial nuclear power reactors because of lack of storage space for spent fuel. Legislation now being considered by Congress includes plans to provide storage space for commercial spent fuel beginning in 1983. Utilities are being encouraged to provide as much storage space as possible in their existing storage facilities, but projections indicate that a significant amount of AFR storage will be required. The government is evaluating the use of both existing and new storage facilities to solve this forecasted storage problem for commercial spent fuel

  17. Ageing management program for reactor components in HANARO

    International Nuclear Information System (INIS)

    Cho, Yeong-Garp; Wu, Sang-Ik; Lee, Jung-Hee; Ryu, Jeong-Soo; Park, Yong-Chul; Wu, Jong-Sup; Jun, Byung Jin

    2003-01-01

    The HANARO, an open-tank-in-pool type research reactor of 30MWth power in Korea, has operated for 8 years since its initial criticality in February of 1995. The reactor power has been gradually increased to 24 MWth through the service period. Therefore the reactor age is very young from the viewpoint of the ageing effect on the reactor structure and components by neutron irradiation considering the expected reactor lifetime. But, we have a few programs to manage the ageing from the aspect of design lifetime of reactor components. This paper summarizes the overall progress and plan for the ageing management for the reactor components including lifetime extension and design improvement, remote measurements and in-service inspections. The shutoff units and control absorber units have aged more rapidly than other structures or components because the number of rod drop cycles was higher than expected at the design stage. The system commissioning tests, periodic performance tests, and weekly operation for the stable supply of medical radioisotopes overriding the normal cycle operation have contributed to the high frequency of rod drop. Therefore, we have instituted a program to extend the lifetime of the shutoff units and the control absorber units. This program includes an endurance test to verify the performance for the extended number of drops and the management of shutdown methods to minimize the drop cycles for both the shutoff units and the control absorber units. The program also includes the design improvement of the damper mechanism of the control absorber units to reduce the impact force caused by rod drop. The inner shell of the reflector vessel surrounding the core is the most critical part from the viewpoint of neutron irradiation. The periodic measurement of the dimensional change in the vertical straightness of the inner shell is considered as one of the in-service inspections. We developed a few tools and verified the performance to measure the

  18. Expert system for control rod programming of boiling water reactors

    International Nuclear Information System (INIS)

    Fukuzaki, T.; Yoshida, K.; Kobayashi, Y.; Matsuura, H.; Hoshi, K.

    1986-01-01

    Control rod programming, one of the main tasks in reactor core management of boiling water reactors (BWRs), can be successfully accomplished by well-experienced engineers. By use of core performance evaluation codes, their knowledge plays the main role in searching through optimal control rod patterns and exposure points for adjusting notch positions and exchanging rod patterns. An expert system has been developed, based on a method of knowledge engineering, to lighten the engineer's load in control rod programming. This system utilizes an inference engine suited for planning/designing problems, and stores the knowledge of well-experienced engineers in its knowledge base. In this report, the inference engine, developed considering the characteristics of the control rod programming, is introduced. Then the constitution and function of the expert system are discussed

  19. A quality assurance program for nuclear power reactor materials tests at the Ford nuclear reactor

    International Nuclear Information System (INIS)

    Burn, R.R.

    1989-01-01

    The University of Michigan Nuclear Reactor Laboratory Quality Assurance Program has been established to assure that materials testing services provided to electric utilities produce accurate results in accordance with industry standards, sound engineering practice, and customer requirements. The program was prepared to comply with applicable requirements of 10CFR50, Appendix B, of the Code of Federal Regulations and a standard of the American National Standards Institute (ANSI), N45.2. The paper discusses the quality assurance program applicability, organization, qualification and training of personnel, material identification and control, examination and testing, measuring and test equipment, nonconforming test equipment, records, audits, and distribution

  20. The reactor engineer program: creating a new workforce

    International Nuclear Information System (INIS)

    Summers, R.

    1993-01-01

    As the number of nuclear engineering schools continues to shrink across the U.S., talented professional engineers for the nuclear energy community must increasingly be found elsewhere. To meet its needs, therefore, the Office of Nuclear Reactor Regulation (NRR) established an Intern Program to bring new talent into the NRC. The two-year program includes 17 weeks of technical training, and 4 or 5 rotational assignments, including at least 4 months at a commercial nuclear power plant site. The key to the success of the program is the full support of NRR high-level management

  1. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  2. SIEX3: A correlated computer code for prediction of fast reactor mixed oxide fuel and blanket pin performance

    International Nuclear Information System (INIS)

    Baker, R.B.; Wilson, D.R.

    1986-04-01

    The SIEX3 computer program was developed to calculate the fuel and cladding performance of oxide fuel and oxide blanket pins irradiated in the fast neutron environment of a liquid metal cooled reactor. The code is uniquely designed to be accurate yet quick running and use a minimum of computer core storage. This was accomplished through the correlation of physically based models to very large data bases of irradiation test results. Data from over 200 fuel pins and over 800 transverse fuel microscopy samples were used in the calibrations

  3. A gold-immobilized microchannel flow reactor for oxidation of alcohols with molecular oxygen.

    Science.gov (United States)

    Wang, Naiwei; Matsumoto, Tsutomu; Ueno, Masaharu; Miyamura, Hiroyuki; Kobayashi, Shū

    2009-01-01

    Golden capillaries: A gold-immobilized capillary column reactor allows oxidation of alcohols to carbonyl compounds using molecular oxygen. These capillary columns (see picture) can be used for at least four days without loss of activity.

  4. Partial oxidation of methane in a temperature-controlled dielectric barrier discharge reactor

    KAUST Repository

    Zhang, Xuming; Cha, Min

    2015-01-01

    We studied the relative importance of the reduced field intensity and the background reaction temperature in the partial oxidation of methane in a temperature-controlled dielectric barrier discharge reactor. We obtained important mechanistic insight

  5. Safety analysis of switching between reductive and oxidative conditions in a reaction coupling reverse flow reactor.

    NARCIS (Netherlands)

    van Sint Annaland, M.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

    2001-01-01

    A new reverse flow reactor is developed where endothermic reactants (propane dehydrogenation) and exothermic reactants (fuel combustion) are fed sequentially to a monolithic catalyst, while periodically alternating the inlet and outlet positions. Upon switching from reductive to oxidative conditions

  6. Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

    International Nuclear Information System (INIS)

    Atkins, D.F.; Lillie, A.F.

    1983-01-01

    As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

  7. The reactor physics computer programs in PC's era

    International Nuclear Information System (INIS)

    Nainer, O.; Serghiuta, D.

    1995-01-01

    The main objective of reactor physics analysis is the evaluation of flux and power distribution over the reactor core. For CANDU reactors sophisticated computer programs, such as FMDP and RFSP, were developed 20 years ago for mainframe computers. These programs were adapted to work on workstations with UNIX or DOS, but they lack a feature that could improve their use and that is 'user friendly'. For using these programs the users need to deal with a great amount of information contained in sophisticated files. To modify a model is a great challenge. First of all, it is necessary to bear in mind all the geometrical dimensions and accordingly, to modify the core model to match the new requirements. All this must be done in a line input file. For a DOS platform, using an average performance PC system, could it be possible: to represent and modify all the geometrical and physical parameters in a meaningful way, on screen, using an intuitive graphic user interface; to reduce the real time elapsed in order to perform complex fuel-management analysis 'at home'; to avoid the rewrite of the mainframe version of the program? The author's answer is a fuel-management computer package operating on PC, 3 time faster than on a CDC-Cyber 830 mainframe one (486DX/33MHz/8MbRAM) or 20 time faster (Pentium-PC), respectively. (author). 5 refs., 1 tab., 5 figs

  8. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    International Nuclear Information System (INIS)

    Karahan, Aydin; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO 2 -PuO 2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  9. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology, MA (United States)

    2010-01-31

    An engineering code to model the irradiation behavior of UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  10. Training and research reactor facility longevity extension program

    International Nuclear Information System (INIS)

    Carriveau, G.W.

    1991-01-01

    Since 1943, over 550 training and research reactors have been in operation. According to statistics from the International Atomic Energy Agency, ∼325 training and research reactors are currently in service. This total includes a wide variety of designs covering a range of power and research capabilities located virtually around the world. A program has been established at General Atomics (GA) that is dedicated to the support of extended longevity of training and research reactor facilities. Aspects of this program include the following: (1) new instrumentation and control systems; (2) improved and upgraded nuclear monitoring and control channels; (3) facility testing, repair and upgrade services that include (a) pool or tank integrity, (b) cooling system, and (c) water purification system; (4) fuel element testing procedures and replacement; (5) control rod drive rebuilding and upgrades; (6) control and monitoring system calibration and repair service; (7) training services, including reactor operations, maintenance, instrumentation calibration, and repair; and (8) expanded or new uses such as neutron radiography and autoradiography, isotope production, nuclear medicine, activation analysis, and material properties modification

  11. The Canadian R and D program targeted at CANDU reactors

    International Nuclear Information System (INIS)

    Moeck, E.O.

    1988-01-01

    CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

  12. Compatibility of sodium with ceramic oxides employed in nuclear reactors; Compatibilidad del sodio con oxidos ceramicos utilizados en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Acena Moreno, V

    1981-07-01

    This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

  13. Computer code qualification program for the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Popov, N.K.; Wren, D.J.; Snell, V.G.; White, A.J.; Boczar, P.G.

    2003-01-01

    Atomic Energy of Canada Ltd (AECL) has developed and implemented 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 provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, and assessment of their applicability in the safety analyses of the ACR design. An outline of the incremental validation program, and an overview of the experimental program in support of the code validation are also presented. An outline of the SQA program used to qualify these computer codes is also briefly presented. To provide context to the differences in the SQA with respect to current CANDUs, the paper also provides an overview of the ACR design features that have an impact on the computer code qualification. (author)

  14. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)

    1999-12-31

    In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

  15. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)

    1998-12-31

    In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

  16. The analysis of reactor vessel surveillance program data

    International Nuclear Information System (INIS)

    Norris, E.B.

    1979-01-01

    Commercial nuclear power reactor vessel surveillance programs are provided by the reactor supplier and are designed to meet the requirements of ASTM Method E 185. (3). Each surveillance capsule contains sets of Charpy V-notch (Csub(v)) specimens representing selected materials from the vessel beltline region and some reference steel, tension test specimens machined from selected beltline materials, temperature monitors, and neutron flux dosimeters. Surveillance capsules may also contain fracture mechanics specimens machined from selected vessel beltline materials. The major steps in the conduct of a surveillance program include (1) the testing of the surveillance specimens to determine the exposure conditions at the capsule location and the resulting embrittlement of the vessel steel, (2) the extrapolation of the capsule results to the pressure vessel wall, and (3) the determination of the heatup and cooldown limits for normal, upset, and test operation. This paper will present data obtained from commercial light water reactor surveillance programs to illustrate the methods of analysis currently in use at Southwest Research Institute and to demonstrate some of the limitations imposed by the data available. Details concerning the procedures for testing the surveillance capsule specimens will not be included because they are considered to be outside of the scope of this paper

  17. The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

    International Nuclear Information System (INIS)

    Townes, B.M.; Hilborn, J.W.

    1985-06-01

    A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

  18. Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

    1991-02-01

    The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

  19. Nuclear data needs for US fast reactor programs

    International Nuclear Information System (INIS)

    Daughtry, J.W.; Rawlins, J.A.

    1985-05-01

    Recent developments in US Fast Reactor Programs are reviewed to provide a background for the nuclear data needs of these programs. Innovative designs, inherent safety, and space nuclear power are receiving increased emphasis. Longstanding and newly-identified nuclear data requirements are reviewed. These requirements are based on information obtained early in 1985 in response to an inquiry sent out by the authors. Finally, plans are outlined for development of an adjusted cross section set for FFTF reload design. The adjustment process suggests some possible changes in ENDF/B-V nuclear data

  20. Preventive maintenance program for a research and production reactor

    International Nuclear Information System (INIS)

    Rico, N.A.

    1990-01-01

    This program proposes a simple, rapid and efficient methodology for the task of developing a really preventive maintenance discipline. Moreover, the lower cost of its application -since it must satisfy the plant's budget-. To this purpose, an extremely economical and easily obtainable infrastructure is proposed. The following stage is referred to the commissioning system, subsequent supervision and follow-up. The experience gained from the two reactors as RA-6 (Bariloche Atomic Center) and NUR (RAE) of Argelia. Finally, the interacting characteristic of this program, since it may be rapidly adapted to different dimensions of plants, laboratories, etc., must be pointed out. (Author) [es

  1. New initiatives in the U.S. Reactor Inspection Program

    International Nuclear Information System (INIS)

    Volgenau, Ernst.

    1977-01-01

    Recently, the United States Nuclear Regulatory Commission (NRC) has initiated a revised inspection approach that will involve placing inspectors full time onsite at all reactor sites. These resident inspectors will be supplemented by a performance appraisal inspection program that will incorporate thorough critical reviews of licensee facilities and an increased program of specific technical measurements to independently verify the accuracy and completeness of licensee work. To complement the inspection initiatives, the NRC is examining ways to expand its enforcement sanctions and to motivate safe licensee performance. (Auth.) [fr

  2. Treatment of spent fuels from research reactors and reactor development programs in Germany

    International Nuclear Information System (INIS)

    Closs, K.D.

    1999-01-01

    Quite a great number of different types of spent fuel from research reactors and development programs exists in Germany. The general policy is to send back to the USA as long as possible fuel from MTRs and TRIGAs of USA origin. An option is reprocessing in Great Britain or France. This option is pursued as long as reprocessing and reuse of the recovered material is economically justifiable. For those fuels which cannot be returned to the USA or which will not be reprocessed, a domestic back-up solution of spent fuel management has been developed in Germany, compatible with the management of spent fuel from power reactors. It consists in dry storage in special casks and, later on, direct disposal. Preliminary results from experimental R and D investigations with research reactor fuel and experience from LWR fuel lead to the conclusion that the direct disposal option even for research reactor fuel or exotic fuel does not impose major technical difficulties for the German waste management and disposal concept. (author)

  3. Planned Scientific programs around the Triga Mark 2 Reactor

    International Nuclear Information System (INIS)

    Majah, M Ibn.

    2007-01-01

    Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

  4. Tubular solid oxide fuel cell development program

    Energy Technology Data Exchange (ETDEWEB)

    Ray, E.R.; Cracraft, C.

    1995-12-31

    This paper presents an overview of the Westinghouse Solid Oxide Fuel Cell (SOFC) development activities and current program status. The Westinghouse goal is to develop a cost effective cell that can operate for 50,000 to 100,000 hours. Progress toward this goal will be discussed and test results presented for multiple single cell tests which have now successfully exceeded 56,000 hours of continuous power operation at temperature. Results of development efforts to reduce cost and increase power output of tubular SOFCs are described.

  5. On-chip microplasma reactors using carbon nanofibres and tungsten oxide nanowires as electrodes

    NARCIS (Netherlands)

    Agiral, A.; Groenland, A.W.; Chinthaginjala, J.K.; Kumar Chinthaginjala, J.; Seshan, Kulathuiyer; Lefferts, Leonardus; Gardeniers, Johannes G.E.

    2008-01-01

    Carbon nanofibres (CNFs) and tungsten oxide (W18O49) nanowires have been incorporated into a continuous flow type microplasma reactor to increase the reactivity and efficiency of the barrier discharge at atmospheric pressure. CNFs and tungsten oxide nanowires were characterized by high-resolution

  6. Optimization geometries of a vortex gliding-arc reactor for partial oxidation of methane

    International Nuclear Information System (INIS)

    Guofeng, Xu; Xinwei, Ding

    2012-01-01

    The effects of the geometry of gliding-arc reactor – such as distance between the electrodes, outlet diameter, and inlet position – on the reactor characteristics (methane conversion, hydrogen yield, and energy efficiency) have not been fully investigated. In this paper, AC gliding-arc reactors including the vortex flow configuration are designed to produce hydrogen from the methane by partial oxidation. The influence of vortex flow configuration on the reactor characteristics is also studied by varying the inlet position. When the inlet of the gliding-arc reactor is positioned close to the outlet, reverse vortex flow reactor (RVFR), the maximum energy efficiency reaches 50% and the yields of hydrogen and carbon monoxide are 40% and 65%, respectively. As the distance between electrodes increases from 5 mm to 15 mm, both hydrogen yield and energy efficiency increase approximately 10% for the RVFR. The energy efficiency and hydrogen yield are highest when the ratio of the outlet diameter to the inner diameter is 0.5 for the RVFR. Experimental results indicate that the flow field in the plasma reactor has an important influence on the reactor performance. Furthermore, hydrogen production increases as the number of feed gas flows in contact with the plasma zone increases. -- Highlights: ► Gliding-arc reactors were designed to produce hydrogen for studying the characteristics of the vortex flow reactor. ► Hydrogen yield of reverse vortex flow reactor was 10% higher than that of forward vortex flow reactor. ► Maximum energy efficiency was 50% for reverse vortex flow reactor. ► If discharge power was supplied to the reactors, the reactor performance increased with increasing distance between electrodes. ► Optimum ratio of the outlet and inner diameter was 1/2.

  7. Assessing the degree of plug flow in oxidation flow reactors (OFRs): a study on a potential aerosol mass (PAM) reactor

    Science.gov (United States)

    Mitroo, Dhruv; Sun, Yujian; Combest, Daniel P.; Kumar, Purushottam; Williams, Brent J.

    2018-03-01

    Oxidation flow reactors (OFRs) have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate). While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs) inside the Washington University Potential Aerosol Mass (WU-PAM) reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS) model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD) simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study).

  8. Assessing the degree of plug flow in oxidation flow reactors (OFRs: a study on a potential aerosol mass (PAM reactor

    Directory of Open Access Journals (Sweden)

    D. Mitroo

    2018-03-01

    Full Text Available Oxidation flow reactors (OFRs have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate. While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs inside the Washington University Potential Aerosol Mass (WU-PAM reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study.

  9. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    Science.gov (United States)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  10. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    International Nuclear Information System (INIS)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  11. Experimental Investigation of Flow Resistance in a Coal Mine Ventilation Air Methane Preheated Catalytic Oxidation Reactor

    OpenAIRE

    Zheng, Bin; Liu, Yongqi; Liu, Ruixiang; Meng, Jian; Mao, Mingming

    2015-01-01

    This paper reports the results of experimental investigation of flow resistance in a coal mine ventilation air methane preheated catalytic oxidation reactor. The experimental system was installed at the Energy Research Institute of Shandong University of Technology. The system has been used to investigate the effects of flow rate (200 Nm3/h to 1000 Nm3/h) and catalytic oxidation bed average temperature (20°C to 560°C) within the preheated catalytic oxidation reactor. The pressure drop and res...

  12. Reactor materials program process water component failure probability

    International Nuclear Information System (INIS)

    Daugherty, W. L.

    1988-01-01

    The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system

  13. Participation in the U.S. Department of Energy Reactor Sharing Program

    International Nuclear Information System (INIS)

    Mulder, R. U.; Benneche, P. E.; Hosticka, B.

    1998-01-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities

  14. Advanced CANDU reactor development: a customer-driven program

    International Nuclear Information System (INIS)

    Hopwood, J.M.

    2005-01-01

    The Advanced CANDU Reactor (ACR) product development program is well under way. The development approach for the ACR is to ensure that all activities supporting readiness for the first ACR project are carded out in parallel, as parts of an integrated whole. In this way design engineering, licensing, development and testing, supply chain planning, construct ability and module strategy, and planning for commissioning and operations, all work in synergy with one another. Careful schedule management :ensures that program focus stays on critical path priorities.'This paper provides an overview of the program, with an emphasis on integration to ensure maximum project readiness, This program management approach is important now that AECL is participating as the reactor vendor in Dominion Energy's DOE-sponsored Combined Construction/Operating License (COL) program. Dominion Energy selected the ACR-700 as their reference reactor technology for purposes of demonstrating the COL process. AECL's development of the ACR is unique in that pre-licensing activities are being carded out parallel in the USA and Canada, via independent, but well-communicated programs. In the short term, these programs are major drivers of ACR development. The ACR design approach has been to optimize to achieve major design objectives: capital cost reduction, robust design with ample margins, proveness by using evolutionary change from existing :reference plants, design for ease :of operability. The ACR development program maintains these design objectives for each of the program elements: Design: .Carefully selected design innovations based on the SEU fuel/light water coolant:/heavy water moderator approach. Emphasis on lessons-learned review from operating experience and customer feedback Licensing: .Safety case based on strengths of existing CANDU plus benefits of optimised design Development and Test: Choice of materials, conditions to enable incremental testing building on existing CANDU and LWR

  15. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

  16. The United States advanced light water reactor (USALWR) development program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Devine, J.C. Jr.; Sugnet, W.R.

    1987-01-01

    For the United States Nuclear Power industry to remain viable, it must be prepared to meet the expected need for a new generation capacity in the late 90s with an improved reactor system. The best hope of meeting this requirement is with revolutionary changes to current LWR systems through simplification and re-evaluation of safety and operational design margins. In addition, the grid characteristics and the difficulty in raising capital for large projects indicate the smaller light water reactors (600 MWe) may play an important role in the next generation. A cooperative and coordinated program between EPRI, U.S. DOE, the major architect engineers, nuclear steam supply vendors, and the NRC in the U.S. has been undertaken with four major goals in mind

  17. The United States Advanced Light Water reactor (USALWR) development program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Devine, Jr.J.C.; Sugnet, W.R.

    1987-01-01

    For the United States Nuclear power industry to remain viable, it must be prepared to meet the expected need for a new generation capacity in the late 90s with an improved reactor system. The best hope of meeting this requirement is with revolutionary changes to current LWR systems through simplification and re-evaluation of safety and operational design margins. In addition, the grid characteristics and the difficulty in raising capital for large projects indicate the smaller light water reactors (600 MWe) may play an important role in the next generation. A cooperative and coordinated program between EPRI, U.S. DOE, the major architect engineers, nuclear steam supply vendors, and the NRC in the U.S. has been undertaken with four major goals in mind. (author)

  18. The US Advanced Liquid-Metal Reactor Program

    International Nuclear Information System (INIS)

    Brolin, E.C.

    1992-01-01

    Based on National Energy Strategy projections, utilities will be required to substantially increase electric generating capacity over the next 40 yr to meet economic growth requirements and replace retiring capacity. Although aggressive conservation measures can save up to 85 GW(electric), ∼195 GW(electric) of additional generating capcity will still be needed by 2010. Assuming startup of new plants around 2000, US Department of Energy (DOE) analyses show that nuclear power can contribute 195 GW(electric) of capacity by 2030, or ∼20% of total electric generation. The DOE is involved in a number of strategies designed to revitalize the nuclear power industry and enable it to meet this projected need for additional capacity. Among these is an integrated overall strategy for advanced reactor development and high-level waste management. A high priority in pursuit of this strategy is the Advanced Liquid-Metal Reactor (ALMR) Program

  19. Japanese Fast Reactor Program for Homogeneous Actinide Recycling

    International Nuclear Information System (INIS)

    Ishikawa, Makoto; Nagata, Takashi; Kondo, Satoru

    2008-01-01

    In the present report, the homogeneous actinide recycling scenario of Fast Reactor (FR) Cycle Technology Development Project (FaCT) is summarized. First, the scenario of nuclear energy policy in Japan are briefly reviewed. Second, the basic plan of Japan to manage all minor actinide (MA) by recycling is summarized objectives of which are the efficiency increase of uranium resources, the environmental burden reduction, and the increase of nuclear non-proliferation potential. Third, recent results of reactor physics study related to MA-loaded FR cores are briefly described. Fourth, typical nuclear design of MA-loaded FR cores in the FaCT project and their main features are demonstrated with the feasibility to recycle all MA in the future FR equilibrium society. Finally, the research and development program to realize the MA recycling in Japan is introduced, including international cooperation projects. (authors)

  20. Program of RA reactor start-up to nominal power; Program dizanja reaktora 'RA' na nominalnu snagu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is desed in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members.

  1. The RERTR [Reduced Enrichment Research and Test Reactor] program:

    International Nuclear Information System (INIS)

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) program is described. After a brief summary of the results which the RERTR program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results and new developments which ocurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40 % average burnup. Good progress was made in the area of LEU usage for the production of fission 99 Mo, and in the coordination of safety evaluations related to LEU conversions of U.S. university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U 3 Si-Al with 19.75 % enrichment and U 3 Si 2 -Al with 45 % enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR program. (Author)

  2. Effect of decontamination on oxidation of austenitic stainless steel in reactor conditions

    International Nuclear Information System (INIS)

    Starkman, T.

    1984-07-01

    Austenitic stainless steels were oxidized in static autoclaves in light water reactor conditions. After the autoclave treatments the specimens were decontaminated with the aid of alkaline potassium permanganate (AP) and oxalic and citric acid (CITROX) as well as electrochemically in H 3 PO 4 . Alternating oxidation and decontamination tests were performed. An elemental analysis of the surfaces of the specimens was carried out by electron spectroscopy. Changes in structures and thicknesses of the oxide layers were observed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

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

  4. SYNBURN: fast-reactor fuel-cycle program

    International Nuclear Information System (INIS)

    Pizzica, P.A.; Meneley, D.A.

    1976-01-01

    The SYNBURN computer program for fast reactors will calculate all the neutronics necessary to completely characterize the equilibrium cycle as well as the startup to equilibrium cycles. The program's run time is very short and this makes the program suitable for survey of parametric studies. It can search on the cycle time for a specified burnup, for the shim control necessary for criticality as well as feed enrichments and the enrichment ratio among core zones. SYNBURN synthesizes in a very simple fashion the one-dimensional fluxes in radial and axial geometry to achieve an approximate two-dimensional solution which agrees very well with the exact two-dimensional solution when measuring regional integrated quantities

  5. German Light-Water-Reactor Safety-Research Program

    International Nuclear Information System (INIS)

    Seipel, H.G.; Lummerzheim, D.; Rittig, D.

    1977-01-01

    The Light-Water-Reactor Safety-Research Program, which is part of the energy program of the Federal Republic of Germany, is presented in this article. The program, for which the Federal Minister of Research and Technology of the Federal Republic of Germany is responsible, is subdivided into the following four main problem areas, which in turn are subdivided into projects: (1) improvement of the operational safety and reliability of systems and components (projects: quality assurance, component safety); (2) analysis of the consequences of accidents (projects: emergency core cooling, containment, external impacts, pressure-vessel failure, core meltdown); (3) analysis of radiation exposure during operation, accident, and decommissioning (project: fission-product transport and radiation exposure); and (4) analysis of the risk created by the operation of nuclear power plants (project: risk and reliability). Various problems, which are included in the above-mentioned projects, are concurrently studied within the Heiss-Dampf Reaktor experiments

  6. Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States); Peko, D. [US Dept. of Energy, Washington, DC (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Humrickhouse, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-06-01

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

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

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

    2015-06-01

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

  8. The RERTR demonstration experiments program at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K; King, J S [Department of Nuclear Engineering, University of Michigan (United States)

    1983-08-01

    The purpose of this paper is to highlight a major part of the experimental work which is being carried out at the Ford Nuclear Reactor (FNR) in conjunction with the RERTR program. A demonstration experiments program has been developed to: 1) characterize the FNR in sufficient detail to discern and quantify neutronic differences between the high and low enriched cores; 2) provide the theoretical group with measurements to benchmark their calculations. As with any experimental program associated with a reactor, stringent constraints limit the experiments which can be performed. Some experiments are performed routinely on the FNR (such as control rod calibrations), and much data is already available. Unfortunately, the accuracy we demand precludes using much of this earlier data. And in many cases, the requirement of precise (and copious) data has led to either developing new techniques (as in the case of rhodium mapping and neutron diffraction) or to further refinements on existing methods (as in the case of spectral unfolding). Nevertheless, we have tried to stay within the realm of recognized, well-established experimental methods in order to assuage any doubts about measured differences between HEU and LEU core parameters. This paper describes the principal results of the experiments performed so far.

  9. Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

    International Nuclear Information System (INIS)

    Broc, Daniel

    2001-01-01

    As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

  10. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  11. Study of Advanced Reactor Mixed Oxide Fuel Production of (U,Th)O2

    International Nuclear Information System (INIS)

    Busron-Masduki; Damunir; Pristi-Hartati; R-Sukarsono; Bangun-Wasito

    2000-01-01

    The high price and starting scarcity of reserved of oil drive the people to drill the alternative nuclear energy. Accelerator-driven Transmutation Waste (ATW) is a prospective technology to solve the problem of used fuel waste, to reduce the anxiety of long term disposal waste, to increase the public acceptance of nuclear energy enter into the third millennium. The future of large nuclear energy appears in many-branched industry will depend on the capability to generate relatively low priced fuel on the basis of commercial nuclear energy. Utilization of uranium-233 -thorium cycle insures long-term fuel supply, makes the nuclear energy production more flexible and enables the self-provision regime to be realized in future. Flowsheet of mixed oxide fuel production for advanced reactor of (U,Th)O 2 is a combination of existing manufacturing equipment and quality assurance program from commercial LWR and HTR. The front-end of flowsheet using sol-gel process. The external sol-gel process is chosen due to simple equipment can anticipate refabrication of U-233 which always contains a few hundred ppm of U-232 and its gamma-emitting daughters, besides yielding smaller waste. The decision to choose external sol-gel process encourages to develop External Gelation Thorium (EGT). In order to get higher density and relatively low compaction pressures (i.e. for advanced LWR) adopted flowsheet EGT is developed to be Sol-Gel Microsphere Pelletization (SGMP). Using the optimal parameters, SGMP become established flowsheet for producing mixed oxide fuel of (U,Th)O 2 for advanced reactor. (author)

  12. Programs with societal benefits at the Cornell University TRIGA reactor

    International Nuclear Information System (INIS)

    Clark, D.D.; Aderhold, H.C.; Hossain, T.Z.

    1993-01-01

    In its 30 yr of operation, the Cornell TRIGA reactor has been used for many educational and research programs that provide general benefits to society. In addition to supporting graduate-level education of nuclear scientists and engineers, it has been extensively used in undergraduate and graduate courses and research by nonspecialists and, through the medium of tours, in education of the general public. Some educational functions have been described previously. In this paper, examples are presented of research of societal interest in nonnuclear fields. The first two rely mainly on radiography, and the remaining five on neutron activation analysis (NAA)

  13. Severe accident sequence assessment for boiling water reactors: program overview

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

  14. The Nuclear Regulatory Commission's research reactor inspection program

    International Nuclear Information System (INIS)

    Constable, George L.

    1977-01-01

    The Division of Reactor Inspection Program's functional responsibilities are presented. The include (a) inspections and investigations necessary to determine whether licensees are complying with license provisions and rules, and to ascertain whether licensed operations are being conducted safely; (b) establishment of bases for the issuance or denial of a construction permit or license; (c) investigation of accidents, incidents, and theft or diversion of special nuclear materials; (d) enforcement actions; and (e) evaluation of licensed operations as a basis for recommending changes to standards and license conditions and for issuance of reports to the nuclear industry and the public

  15. Experimental Investigation of Flow Resistance in a Coal Mine Ventilation Air Methane Preheated Catalytic Oxidation Reactor

    Directory of Open Access Journals (Sweden)

    Bin Zheng

    2015-01-01

    Full Text Available This paper reports the results of experimental investigation of flow resistance in a coal mine ventilation air methane preheated catalytic oxidation reactor. The experimental system was installed at the Energy Research Institute of Shandong University of Technology. The system has been used to investigate the effects of flow rate (200 Nm3/h to 1000 Nm3/h and catalytic oxidation bed average temperature (20°C to 560°C within the preheated catalytic oxidation reactor. The pressure drop and resistance proportion of catalytic oxidation bed, the heat exchanger preheating section, and the heat exchanger flue gas section were measured. In addition, based on a large number of experimental data, the empirical equations of flow resistance are obtained by the least square method. It can also be used in deriving much needed data for preheated catalytic oxidation designs when employed in industry.

  16. The burnup dependence of light water reactor spent fuel oxidation

    International Nuclear Information System (INIS)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO 2 is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO 2 to higher oxides. The oxidation of UO 2 has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO 2 oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO 2 to UO 2.4 was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO 2.4 to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO 2 oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO 2 and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5)

  17. Technological studies for obtaining lead oxide compacts used in generation IV nuclear reactors

    International Nuclear Information System (INIS)

    Paraschiv, I.; Benga, D.

    2016-01-01

    One of the main concerns of the nuclear research at this moment is the development of the necessary technologies for Generation IV reactors. The main candidate as coolant agent in these reactors is molten lead but this material involves ensuring the oxygen control, due to potential contamination of coolant through the formation of solid oxides and the influence on the corrosion rate of structural parts and for this reason, the oxygen concentration must be kept in a well specified domain. One of the proposed methods for oxygen monitoring and control in the technology of Generation IV reactors, is the use of PbO compacts. For this paper technological tests were performed for developing and setting the optimal parameters in order to attain lead oxide compacts necessary for the oxygen control technology in Generation IV nuclear reactors. (authors)

  18. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1992-01-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  19. A review of fast reactor program in Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  20. A new oxidation flow reactor for measuring secondary aerosol formation of rapidly changing emission sources

    Science.gov (United States)

    Simonen, Pauli; Saukko, Erkka; Karjalainen, Panu; Timonen, Hilkka; Bloss, Matthew; Aakko-Saksa, Päivi; Rönkkö, Topi; Keskinen, Jorma; Dal Maso, Miikka

    2017-04-01

    Oxidation flow reactors (OFRs) or environmental chambers can be used to estimate secondary aerosol formation potential of different emission sources. Emissions from anthropogenic sources, such as vehicles, often vary on short timescales. For example, to identify the vehicle driving conditions that lead to high potential secondary aerosol emissions, rapid oxidation of exhaust is needed. However, the residence times in environmental chambers and in most oxidation flow reactors are too long to study these transient effects ( ˜ 100 s in flow reactors and several hours in environmental chambers). Here, we present a new oxidation flow reactor, TSAR (TUT Secondary Aerosol Reactor), which has a short residence time ( ˜ 40 s) and near-laminar flow conditions. These improvements are achieved by reducing the reactor radius and volume. This allows studying, for example, the effect of vehicle driving conditions on the secondary aerosol formation potential of the exhaust. We show that the flow pattern in TSAR is nearly laminar and particle losses are negligible. The secondary organic aerosol (SOA) produced in TSAR has a similar mass spectrum to the SOA produced in the state-of-the-art reactor, PAM (potential aerosol mass). Both reactors produce the same amount of mass, but TSAR has a higher time resolution. We also show that TSAR is capable of measuring the secondary aerosol formation potential of a vehicle during a transient driving cycle and that the fast response of TSAR reveals how different driving conditions affect the amount of formed secondary aerosol. Thus, TSAR can be used to study rapidly changing emission sources, especially the vehicular emissions during transient driving.

  1. The United States foreign research reactor spent nuclear fuel acceptance program: Proposal to modify the program

    International Nuclear Information System (INIS)

    Messick, C.E.

    2005-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. The policy was slated to expire in May 2009. However, in October 2003, a petition requesting a program extension was delivered to the United States Secretary of Energy from a group of research reactor operators from foreign countries. In April 2004, the Secretary directed DOE undertake an analysis, as required by the National Environmental Policy Act (NEPA), to consider potential extension of the Program. On December 1, 2004, a Federal Register Notice was issued approving the program extension. This paper discusses the findings from the NEPA analysis and the potential changes in the program that may result from implementation of the proposed changes. (author)

  2. Quality assurance program plan for the Reactor Research Experiment Programs (RREP)

    International Nuclear Information System (INIS)

    Pipher, D.G.

    1982-05-01

    This document describes the Quality Assurance Program plans which will be applied to tasks on Reactor Research Experiments performed on Sandia National Laboratories' reactors. The program provides for individual project or experiment quality plan development and allows for reasonable plan flexibility and maximum plan visibility. Various controls and requirements in this program plan are considered mandatory on all features which are identified as important to public health and safety (Level I). It is the intent of this document that the Quality Assurance program comprise those elements which will provide adequate assurance that all components, equipment, and systems of the experiments will perform as designed, and hence prevent delays and costs due to rejections or failures

  3. Research program and uses of the solution fueled reactor SILENE

    International Nuclear Information System (INIS)

    Barbry, F.; Ratel, R.

    1985-09-01

    Designed and operated by the Nuclear Protection and Safety Institute of the CEA, SILENE is an original small sized reactor fueled with an uranyl nitrate solution. The reactor is capable to operate in three modes: ''Pulse'' operation (high power levels up to 1000 Megawatts during several millisecond), ''Free evolution'' operation (simulation of criticality accident excursions), ''Steady state'' operation in a power range of 0.01 W to 1 kW. The core can be surrounded by appropriate shields (lead, polyethylene) to vary the leakage radiations and the gamma to neutron dose ratio. It's possible to insert in the central cavity of the annular core vessel some capsules, devices or samples to be submitted to very high radiations levels. The research activities are mainly devoted towards nuclear safety studies: the criticality accident studies, and the behavior of oxide fuels under transient conditions. Some examples of tests are presented. As to other applications of the SILENE facility, the main studies now in progress deal with: designing and calibration of Health physics intrumentation, neutron and gamma dosimetry, and, radiobiology. Once the characteristics of radiation field are qualified by calculations and experimental techniques, SILENE will be proposed as a reference source [fr

  4. Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment

    Energy Technology Data Exchange (ETDEWEB)

    Quastel, A.; Thiriet, C. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Lewis, B., E-mail: brent.lewis@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada); Corcoran, E., E-mail: emily.corcoran@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2014-07-01

    An experimental program funded by the CANDU Owners Group (COG) is studying an out-reactor instrumented defected fuel experiment in Stern Laboratories (Hamilton, Ontario) with guidance from Atomic Energy of Canada Limited (AECL). The objective of this test is to provide experimental data for validation of a mechanistic fuel oxidation model. In this experiment a defected fuel element with UO{sub 2} pellets will be internally heated with an electrical heater element, causing the fuel to crack. By defecting the sheath in-situ the fuel will be exposed to light water coolant near normal reactor operating conditions (pressure 10 MPa and temperature 265-310{sup o}C) causing fuel oxidation, especially near the hotter regions of the fuel in the cracks. The fuel thermal conductivity will change, resulting in a change in the temperature distribution of the fuel element. This paper provides 2D r-θ plane strain solid mechanics models to simulate fuel thermal expansion, where conditions for fuel crack propagation are investigated with the thermal J integral to predict fuel crack stress intensity factors. Finally since fuel crack geometry can affect fuel oxidation this paper shows that the solid mechanics model with pre-set radial cracks can be coupled to a 2D r-θ fuel oxidation model. (author)

  5. Experience in reactor research and development programs as educational system for thermohydraulic engineering

    International Nuclear Information System (INIS)

    Zaki, G.M.; Fikry, M.M.

    1977-01-01

    A reactor development program within a research reactor facility can be used for personnel training on the operation of power reactors and research in the different fields of nuclear science and engineering. A training program is proposed where reactor maintenance and operation, in addition to conducting development programs and executing projects, are utilized for forming specialized groups. The paper gives a short survey of a heat transfer program where out of pile and in-core studies are conducted along with two-phase flow investigations. This program covers the main requirements for WWR (water cooled and moderated reactor) power uprating and furnishes basic knowledge on power reactor thermal parameters. The major facilities for conducting similar programs devoted to education are mentioned

  6. Temperature oscillations in methanol partial oxidation reactor for the production of hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jinsu; Byeon, Jeonguk; Seo, Il Gyu; Lee, Hyun Chan; Kim, Dong Hyun; Lee, Jietae [Kyungpook National University, Daegu (Korea, Republic of)

    2013-04-15

    Methanol partial oxidation (POX) is a well-known reforming reaction for the production of hydrogen from methanol. Since POX is relatively fast and highly exothermic, this reforming method will be efficient for the fast start-up and load-following operation. However, POX generates hot spots around catalyst and even oscillations in the reactor temperature. These should be relieved for longer operations of the reactor without catalyst degradations. For this, temperature oscillations in a POX reactor are investigated experimentally. Various patterns of temperature oscillations according to feed flow rates of reactants and reactor temperatures are obtained. The bifurcation phenomena from regular oscillations to chaotic oscillations are found as the methanol flow rate increases. These experimental results can be used for theoretical analyses of oscillations and for designing safe reforming reactors.

  7. Temperature oscillations in methanol partial oxidation reactor for the production of hydrogen

    International Nuclear Information System (INIS)

    Kim, Jinsu; Byeon, Jeonguk; Seo, Il Gyu; Lee, Hyun Chan; Kim, Dong Hyun; Lee, Jietae

    2013-01-01

    Methanol partial oxidation (POX) is a well-known reforming reaction for the production of hydrogen from methanol. Since POX is relatively fast and highly exothermic, this reforming method will be efficient for the fast start-up and load-following operation. However, POX generates hot spots around catalyst and even oscillations in the reactor temperature. These should be relieved for longer operations of the reactor without catalyst degradations. For this, temperature oscillations in a POX reactor are investigated experimentally. Various patterns of temperature oscillations according to feed flow rates of reactants and reactor temperatures are obtained. The bifurcation phenomena from regular oscillations to chaotic oscillations are found as the methanol flow rate increases. These experimental results can be used for theoretical analyses of oscillations and for designing safe reforming reactors

  8. Status of reduced enrichment program for research reactors in Japan

    International Nuclear Information System (INIS)

    Kaieda, Keisuke; Baba, Osamu; Nagaoka, Yoshiharu; Kanda, Keiji; Nakagome, Yoshihiro

    1999-01-01

    The reduced enrichment programs for the JRR-3M, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI) have been completed. The KUR of Kyoto University Research Reactor Institute (KURRI) has been partially completed and is still in progress under the Joint Study Program with Argonne National Laboratory (ANL). The JRR-3M commenced using LEU silicide fuel elements instead of LEU aluminide fuel elements in September, 1999. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and April 1994 the U.S. Government gave an approval to utilize HEU fuel in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until March 2004, then the full core conversion with LEU silicide will be done. The first shipment of spent fuels since 1974 was done in August, 1999. (author)

  9. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    International Nuclear Information System (INIS)

    Stephens, S.V.

    1976-09-01

    The objective of the document is to present in one volume an overview of the Argonne National Laboratory test facilities involved in the conduct of the national LMFBR research and development program. Existing facilities and those under construction or authorized as of September 1976 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. The volume is divided into two sections: Argonne-East and Argonne-West. Introductory material for each section includes site and facility maps. The profiles are arranged alphabetically by title according to their respective locations at Argonne-East or Argonne-West. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc., involved in the LMFBR program is appended

  10. Prioritization of R and D programs on probabilistic reactor safety

    International Nuclear Information System (INIS)

    Husseiny, A.A.

    1982-01-01

    An interactive computer code based on the multiattribute utility theory has been developed with graphic capabilities to use in selection of probabilistic reactor safety RandD programs. Utility values and proper graphic representation are made through lottery games on the computer terminal. The code is applied to prioritize a set of RandD programs on LWR safety based on attributes including regulatory issues, institutional issues and operation problems. The methodology is described here in detail with its applications. Some of the input includes statistical distributions and subjective judgments on institutional issues. The flexibility of the approach provides a tool for decision makers whether on individual or group level to assess LWR safety priorities and continuously update their strategies

  11. Program for studying fundamental interactions at the PIK reactor facilities

    International Nuclear Information System (INIS)

    Serebrov, A. P.; Vassiljev, A. V.; Varlamov, V. E.; Geltenbort, P.; Gridnev, K. A.; Dmitriev, S. P.; Dovator, N. A.; Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A.; Martemyanov, V. P.

    2016-01-01

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4′ channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4′ channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  12. The Program Planned for the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Haubenreich, Paul N.

    1967-01-01

    This document outlines the program planned for the MSRE in fiscal years 1968 and 1969. It includes a bar diagram of the program, a critical-path type diagram of the operations, and a brief description of each task. In addition to the work at the reactor site, the outline also covers activities elsewhere at ORNL and by the AEC that directly affect the reactor schedule. The amount of detail and the accuracy with which we can estimate times varies considerably among the different items on the schedule. Some items, such as annual checkouts and core sample replacement, have been done before and our time estimates do not include any contingency, In the case of such tasks as planning, reviewing, and preparing for experiments or operations, we have set target dates that appear reasonable and we fully expect to meet these. Processing the salt is a different matter. If there are no unforeseen difficulties we should finish easily in the time shown, but the operation is in part a shakedown, so delays would not be too surprising, The time for modifying the system and adding fluoroborate is, of course, uncertain because the requirements are not yet known. As the requirements develop in more detail the estimate will be updated, but we do not foresee any major dislocation in the schedule, The scheduled time for preparation of enriching salt is becoming tight because of delays in facility construction. Should there be further delays in this key item, the entire schedule would have to be reconsidered.

  13. Program for studying fundamental interactions at the PIK reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Vassiljev, A. V.; Varlamov, V. E. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation); Geltenbort, P. [Institut Laue-Langevin (France); Gridnev, K. A. [St. Petersburg State University (Russian Federation); Dmitriev, S. P.; Dovator, N. A. [Russian Academy of Sciences, Ioffe Physical-Technical Institute (Russian Federation); Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P. [National Research Center Kurchatov Institute (Russian Federation); and others

    2016-05-15

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4′ channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4′ channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  14. Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

    KAUST Repository

    Bin Hamzah, Muhamad Firdaus

    2017-01-01

    Ignition and combustion characteristics of n-butanol/air, 2-butanol.air and isobutanol/air mixtures at stoichiometric (ϕ = 1) and lean (ϕ = 0.5) conditions were investigated in a micro flow reactor with a controlled temperature profile from 323 K

  15. Progress and status of the integral fast reactor (IFR) development program

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    This paper discusses the Integral Fast Reactor (IFR) development program, in which the entire reactor system - reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are also presented

  16. Summary of ORNL high-temperature gas-cooled reactor program

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-01-01

    Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

  17. Electrochemical Oxidation of Propene with a LSF15/CGO10 Electrochemical Reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2014-01-01

    A porous electrochemical reactor, made of La0.85Sr0.15FeO3 (LSF) as electrode and Ce0.9Gd0.1O1.95 (CGO) as electrolyte, was studied for the electrochemical oxidation of propene over a wide range of temperatures. Polarization was found to enhance propene oxidation rate. Ce0.9Gd0.1O1.95 was used...... as infiltration material to enhance the effect of polarization on propene oxidation rate, especially at low temperatures. The influence of infiltrated material, as a function of heat treatment, on the reactor electrochemical behavior has been evaluated by using electrochemical impedance spectroscopy...... in suppressing the competing oxygen evolution reaction and promoting the oxidation of propene under polarization, with faradaic efficiencies above 70% at 250◦C. © 2014 The Electrochemical Society....

  18. Fast reactors with axial arrangement of oxide and metal fuels in the core

    International Nuclear Information System (INIS)

    Troyanov, M.F.; Ilyunin, V.G.; Matveev, V.I.; Murogov, V.M.; Proshkin, A.A.; Rudneva, V.Ya.; Shmelev, A.N.

    1980-01-01

    Problems of using metal fuel in fast reactor (FR) core are discussed Results are given of the calculation of two-dimentional (R-Z) FR version having a composed core with the combined usage of oxide and metal fuels having parameters close to optimal from the point of view of fuel breeding rate, an oxide subzone having increased enrichment and a decreased proper conversion ratio. A reactor is considered where metallic fuel elements are placed from the side of ''cold'' coolant inlet (400-480 deg C), and oxide fuel elements - in the region where the coolant has a higher temperature (500-560 deg C). It is shown that the new fuel breeding rate in such a reactor can be increased by 20-30% as compared with an oxide fuel reactor. Growth of the total conversion ratio is mainly stipulated with the increase of the inner conversion ratio of the core (CRC) which is important not only from the point of view of nuclear fuel breeding rate but also the optimization of the mode of powerful fast reactor operation with provision for the change in reactivity in the process of its continuous operation. The fact, that the core version under investigation has a CRC value slightly exceeding unit, stipulates considerably less reactivity change as compared with the oxide version in the process of the reactor operation and permits at a constant reactor control system power to significantly increase the time between reloadings and, therefore, to increase the NPP load factor which is of great importance both from the point of view of economy and the improvement of operation conditions as well as of reactor operation reliability. It is concluded on the base of the analysis of the results obtained that FRs with the combined usage of oxide and metal fuels having an increased specific load and increased conversion ratio as compared with the oxide fuel FRs provide a higher rate of development of the whole nuclear power balanced with respect to the fuel [ru

  19. U.S. fast reactor materials and structures program

    International Nuclear Information System (INIS)

    Harms, W.O.; Purdy, C.M.

    1984-01-01

    The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program

  20. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  1. On the chemical constitution of a molten oxide core of a fast breeder reactor

    International Nuclear Information System (INIS)

    Hodkin, D.J.; Potter, P.E.

    1980-01-01

    A knowledge of the chemical constitution of a molten oxide fast reactor core is of great importance in the assessment of heat transfer from a cooling molten pool of debris and in the selection of materials for the construction of sacrificial beds for core containment. In this paper we describe some thermodynamic assessments of the likely chemical constitution of a molten oxide core, and then support our assessments by experimental observations

  2. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  3. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  4. Progress report on reactor physics research program, January 1963 - February 1964

    International Nuclear Information System (INIS)

    1964-02-01

    This progress report is a part of the annual report of the department of reactor physics prepared for the Boris Kidric Institute of nuclear sciences. It is a review of research activities in the field of theoretical and experimental reactor physics in the year 1973. A part of this program was included in the NPY Cooperative program in reactor physics. The topics covered by this report are as follows: Calculations of the thermal neutron distribution and reaction rate in a reactor cell and comparison with experiments; buckling measurements; thermalization and slowing down of neutrons; pulsed neutron source techniques; and reactor kinetics

  5. Progress report on reactor physics research program, January 1963 - February 1964

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-02-15

    This progress report is a part of the annual report of the department of reactor physics prepared for the Boris Kidric Institute of nuclear sciences. It is a review of research activities in the field of theoretical and experimental reactor physics in the year 1973. A part of this program was included in the NPY Cooperative program in reactor physics. The topics covered by this report are as follows: Calculations of the thermal neutron distribution and reaction rate in a reactor cell and comparison with experiments; buckling measurements; thermalization and slowing down of neutrons; pulsed neutron source techniques; and reactor kinetics.

  6. Nuclear reactor safety program in US department of energy and future perspectives

    International Nuclear Information System (INIS)

    Song, Y.T.

    1988-01-01

    The US Department of Energy (DOE) establishes policy, issues orders, and assures compliance with requirements. The contractors who design, construct, modify, operate, maintain and decommission DOE reactors, set forth the assessment of the safety of cognizant reactors and implement DOE orders. Teams of experts in the Department, through scheduled and unscheduled review programs, reassess the safety of reactors in every phases of their lives. As new technology develops, the safety programs are reevaluated and policies are modified to accommodate these new technologies. The diagnostic capabilities of the computer using multiple alarms to enhance detection of defects and control of a reactor have been greatly utilized in reactor operating systems. The Application of artificial intelligence technologies for diagnostic and even for the decision making process in the event of reactor accidents would be one of the future trends in reactor safety programs

  7. Nuclear reactor safety program in U.S. Department of Energy and future perspectives

    International Nuclear Information System (INIS)

    Song, Y.T.

    1987-01-01

    The U.S. Department of Energy (DOE) establishes policy, issues orders, and assures compliance with requirements. The contractors who design, construct, modify, operate, maintain and decommission DOE reactors, set forth the assessment of the safety of cognizant reactors and impliment DOE orders. Teams of experts in the Depatment, through scheduled and unscheduled review programs, reassess the safety of reactors in every phases of their lives. As new technology develops, the safety programs are reevaluated and policies are modified to accommodate these new technologies. The diagnostic capabilities of the computer using multiple alarms to enhance detection of defects and control of a reactor have been greatly utilized in reactor operating systems. The application of artificial intelligence (AI) technologies for diagnostic and even for the decision making process in the event of reactor accidents would be one of the future trends in reactor safety programs. (author)

  8. TREATMENT OF REFRACTORY OXIDES IN HF-PLASMA REACTORS

    OpenAIRE

    Bakhvalov , A.; Dresvin , S.; Levitskaya , T.; Paskalov , G.; Philippov , A.

    1990-01-01

    Results of theoretical and experimental studies of SiO2 NaBSi, MgO, W and some other materials treatment in induction type high-frequency plasma under atmospheric pressure are presented. Key study objective - optimization of plasma installation operating modes with maximum efficiency -0.6 -0.7 ; spheroidization extent -90-99%, size of treated particles 1-500 mkm. Diagnostics of thermophysical and gasodynamical plasma reactor specifications has been presented.

  9. DOE University Reactor Sharing Program. Final technical report for 1996--1997

    International Nuclear Information System (INIS)

    Chappas, W.J.; Adams, V.G.

    1998-01-01

    The Department of Energy University Reactor Sharing Program at University of Maryland, College Park (UMCP) has, once again, stimulated a broad use of the reactor and radiation facilities by undergraduate and graduate students, visitors, and professionals. Participants are exposed to topics such as nuclear engineering, radiation safety, and nuclear reactor operations. This information is presented through various means including tours, slide presentations, experiments, and discussions. Student research using the MUTR is also encouraged. In addition, the Reactor Sharing Program here at the University of Maryland does not limit itself to the confines of the TRIGA reactor facility. Incorporated in the program are the Maryland University Radiation Effects Laboratory, and the UMCP 2 x 4 Thermal Hydraulic Loop. These facilities enhance and give an added dimension to the tours and experiments. The Maryland University Training Reactor (MUTR) and the associated laboratories are made available to any interested institution six days a week on a scheduled basis. Most institutions are scheduled at the time of their first request--a reflection of their commitment to the Reactor Sharing Program. The success of the past years by no means guarantees future success. Therefore, the reactor staff is more aggressively pursuing its outreach program, especially with junior colleges and universities without reactor or radiation facilities; more aggressively developing demonstration and training programs for students interested in careers in nuclear power and radiation technology; and more aggressively up-grading the reactor facilities--not only to provide a better training facility but to prepare for relicensing in the year 2000

  10. Identification of significant process variables for a flow-through supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Rossi, R.E.

    1992-05-01

    The effects of four process variables on the destruction efficiency of a flow-through supercritical water oxidation reactor were investigated. These process variables included: (1) reactor throughput (GPH), (2) concentration of the surrogate waste (% acetone), (3) maximum reactor tube-wall temperature (OC), and (4) applied stoichiometric oxygen. The analysis was conducted utilizing two-level factorial experiments, steepest ascent methods, and central composite designs. This experimental protocol assures efficient experimentation and allows for an empirical response surface model of the system to be developed. This experimentation identified a significant positive effect for stoichiometric oxygen applied and temperature variations between 400 to 500 degrees C. The increase in destruction efficiency due to stoichiometric 0 2 provides strong evidence that supercritical water oxidations are catalyzed by excess oxygen, and the strong temperature effect is a result of large increases in the kinetic rates for this temperature range. However, increasing temperature between 550 to 650 degrees C does not provide substantial increases in destruction efficiency. In addition, destruction efficiency is significantly unproved by increasing the Reynolds number and residence time. The destruction efficiency of the reactor is also dependent upon the initial concentration of surrogate waste. This concentration dependence may indicate first-order supercritical CO kinetics is inadequate for describing all waste types and reactor configurations. Alternatively, it may indicate reactant mixing, caused by local turbulence at the oxidation fronts of these higher concentration waste streams, results in higher destruction efficiencies

  11. MIT nuclear reactor laboratory high school teaching program

    International Nuclear Information System (INIS)

    Olmez, I.

    1991-01-01

    For the last 6 years, the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory's academic and scientific staff a have been conducting evening seminars for precollege science teachers, parents, and high school students from the New England area. These seminars, as outlined in this paper, are intended to give general information on nuclear technologies with specific emphasis on radiation physics, nuclear medicine, nuclear chemistry, and ongoing research activities at the MIT research reactor. The ultimate goal is to create interest or build on the already existing interest in science and technology by, for example, special student projects. Several small projects have already been completed ranging from environmental research to biological reactions with direct student involvement. Another outcome of these seminars was the change in attitudes of science teachers toward nuclear technology. Numerous letters have been received from the teachers and parents stating their previous lack of knowledge on the beneficial aspects of nuclear technologies and the subsequent inclusion of programs in their curriculum for educating students so that they may also develop a more positive attitude toward nuclear power

  12. PtRu colloid nanoparticles for CO oxidation in microfabricated reactors

    DEFF Research Database (Denmark)

    Klerke, Asbjørn; Saadi, Souheil; Toftegaard, Maja Bøg

    2006-01-01

    The catalytic activity of PtRu colloid nanoparticles for CO oxidation is investigated in microfabricated reactors. The measured catalytic performance describes a volcano curve as a function of the Pt/Ru ratio. The apparent activation energies for the different alloy catalysts are between 21 and 1...

  13. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2013-12-15

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m{sup 2} s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface.

  14. Oxidation of ethene in a wall-cooled packed-bed reactor

    NARCIS (Netherlands)

    Schouten, E.P.S.; Borman, P.C.; Westerterp, K.R.

    1994-01-01

    The selective oxidation of ethene over a silver on α-alumina catalyst was studied in a pilot plant with a wall-cooled tubular packed bed reactor. Gas and solid temperatures in the catalyst bed were measured at different axial and radial positions as well as concentrations at different axial

  15. TRICHLOROETHYLENE SORPTION AND OXIDATION USING A DUAL FUNCTION SORBENT/CATALYST IN A FALLING FURNACE REACTOR

    Science.gov (United States)

    A dual function medium (Cr-ZSM-5), capable of physisorbing trichloroethylene (TCE) at ambient temperature and catalytically oxidizing it at elevated temperature (-350 degrees C) was utilized in a novel continuous falling furnace reactor system to store and periodically destroy t...

  16. Development of packed bed membrane reactor for the oxidative dehydrogenation of propane

    NARCIS (Netherlands)

    Kotanjac, Zeljko

    2009-01-01

    In this research, a reactor concept for the oxidative dehydrogenation of propane was studied. First a literature survey was performed, to investigate which are the best catalyst systems and best operating conditions that result in the largest propylene yield. In the kinetic study of ODHP over a

  17. Comparison of open cycles of uranium and mixed oxides of thorium-uranium using advanced reactors

    International Nuclear Information System (INIS)

    Gonçalves, Letícia C.; Maiorino, José R.

    2017-01-01

    A comparative study of the mass balance and production costs of uranium oxide fuels was carried out for an AP1000 reactor and thorium-uranium mixed oxide in a reactor proposal using thorium called AP-Th1000. Assuming the input mass values for a fuel load the average enrichment for both reactors as well as their feed mass was determined. With these parameters, the costs were calculated in each fuel preparation process, assuming the prices provided by the World Nuclear Association. The total fuel costs for the two reactors were quantitatively compared with 18-month open cycle. Considering enrichment of 20% for the open cycle of mixed U-Th oxide fuel, the total uranium consumption of this option was 50% higher and the cost due to the enrichment was 70% higher. The results show that the use of U-Th mixed oxide fuels can be advantageous considering sustainability issues. In this case other parameters and conditions should be investigated, especially those related to fuel recycling, spent fuel storage and reduction of the amount of transuranic radioactive waste

  18. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    International Nuclear Information System (INIS)

    Park, Seong Dae; Bang, In Cheol

    2013-01-01

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m 2 s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface

  19. TERRAPOWER, LLC TRAVELING WAVE REACTOR DEVELOPMENT PROGRAM OVERVIEW

    Directory of Open Access Journals (Sweden)

    PAVEL HEJZLAR

    2013-11-01

    burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

  20. Controlled nitric oxide production via O(1D) + N2O reactions for use in oxidation flow reactor studies

    Science.gov (United States)

    Lambe, Andrew; Massoli, Paola; Zhang, Xuan; Canagaratna, Manjula; Nowak, John; Daube, Conner; Yan, Chao; Nie, Wei; Onasch, Timothy; Jayne, John; Kolb, Charles; Davidovits, Paul; Worsnop, Douglas; Brune, William

    2017-06-01

    Oxidation flow reactors that use low-pressure mercury lamps to produce hydroxyl (OH) radicals are an emerging technique for studying the oxidative aging of organic aerosols. Here, ozone (O3) is photolyzed at 254 nm to produce O(1D) radicals, which react with water vapor to produce OH. However, the need to use parts-per-million levels of O3 hinders the ability of oxidation flow reactors to simulate NOx-dependent secondary organic aerosol (SOA) formation pathways. Simple addition of nitric oxide (NO) results in fast conversion of NOx (NO + NO2) to nitric acid (HNO3), making it impossible to sustain NOx at levels that are sufficient to compete with hydroperoxy (HO2) radicals as a sink for organic peroxy (RO2) radicals. We developed a new method that is well suited to the characterization of NOx-dependent SOA formation pathways in oxidation flow reactors. NO and NO2 are produced via the reaction O(1D) + N2O → 2NO, followed by the reaction NO + O3 → NO2 + O2. Laboratory measurements coupled with photochemical model simulations suggest that O(1D) + N2O reactions can be used to systematically vary the relative branching ratio of RO2 + NO reactions relative to RO2 + HO2 and/or RO2 + RO2 reactions over a range of conditions relevant to atmospheric SOA formation. We demonstrate proof of concept using high-resolution time-of-flight chemical ionization mass spectrometer (HR-ToF-CIMS) measurements with nitrate (NO3-) reagent ion to detect gas-phase oxidation products of isoprene and α-pinene previously observed in NOx-influenced environments and in laboratory chamber experiments.

  1. Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.

    2008-01-01

    Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

  2. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  3. A fast-running fuel management program for a CANDU reactor

    International Nuclear Information System (INIS)

    Choi, Hangbok

    2000-01-01

    A fast-running fuel management program for a CANDU reactor has been developed. The basic principle of this program is to select refueling channels such that the reference reactor conditions are maintained by applying several constraints and criteria when selecting refueling channels. The constraints used in this program are the channel and bundle power and the fuel burnup. The final selection of the refueling channel is determined based on the priority of candidate channels, which enhances the reactor power distribution close to the time-average model. The refueling simulation was performed for a natural uranium CANDU reactor and the results were satisfactory

  4. Planar ceramic membrane assembly and oxidation reactor system

    Science.gov (United States)

    Carolan, Michael Francis; Dyer, legal representative, Kathryn Beverly; Wilson, Merrill Anderson; Ohm, Ted R.; Kneidel, Kurt E.; Peterson, David; Chen, Christopher M.; Rackers, Keith Gerard; Dyer, deceased, Paul Nigel

    2007-10-09

    Planar ceramic membrane assembly comprising a dense layer of mixed-conducting multi-component metal oxide material, wherein the dense layer has a first side and a second side, a porous layer of mixed-conducting multi-component metal oxide material in contact with the first side of the dense layer, and a ceramic channeled support layer in contact with the second side of the dense layer. The planar ceramic membrane assembly can be used in a ceramic wafer assembly comprising a planar ceramic channeled support layer having a first side and a second side; a first dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the first side of the ceramic channeled support layer; a first outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the first dense layer; a second dense layer of mixed-conducting multi-component metal oxide material having an inner side and an outer side, wherein the inner side is in contact with the second side of the ceramic channeled layer; and a second outer support layer comprising porous mixed-conducting multi-component metal oxide material and having an inner side and an outer side, wherein the inner side is in contact with the outer side of the second dense layer.

  5. Improving Corrosion Behavior in SCWR, LFR and VHTR Reactor Materials by Formation of a Stable Oxide

    International Nuclear Information System (INIS)

    Motta, Arthur T.; Comstock, Robert; Li, Ning; Allen, Todd; Was, Gary

    2009-01-01

    The objective of this study is to understand the influence of the alloy microstructure and composition on the formation of a stable, protective oxide in the environments relevant to the SCWR and LFR reactor concepts, as well as to the VHTR. It is proposed to use state-of-the art techniques to study the fine structure of these oxides to identify the structural differences between stable and unstable oxide layers. The techniques to be used are microbeam synchrotron radiation diffraction and fluorescence, and cross-sectional transmission electron microcopy on samples prepared using focused ion beam.

  6. Assessing the antimicrobial activity of zinc oxide thin films using disk diffusion and biofilm reactor

    International Nuclear Information System (INIS)

    Gittard, Shaun D.; Perfect, John R.; Monteiro-Riviere, Nancy A.; Wei Wei; Jin Chunming; Narayan, Roger J.

    2009-01-01

    The electronic and chemical properties of semiconductor materials may be useful in preventing growth of microorganisms. In this article, in vitro methods for assessing microbial growth on semiconductor materials will be presented. The structural and biological properties of silicon wafers coated with zinc oxide thin films were evaluated using atomic force microscopy, X-ray photoelectron spectroscopy, and MTT viability assay. The antimicrobial properties of zinc oxide thin films were established using disk diffusion and CDC Biofilm Reactor studies. Our results suggest that zinc oxide and other semiconductor materials may play a leading role in providing antimicrobial functionality to the next-generation medical devices

  7. Advanced Light Water Reactor Program: Program management and staff review methodology

    International Nuclear Information System (INIS)

    Moran, D.H.

    1986-12-01

    This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond

  8. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

  9. Oxidation of carbon based material for innovative energy systems (HTR, fusion reactor): status and further needs

    International Nuclear Information System (INIS)

    Moormann, R.; Hinssen, H.K.; Latge, Ch.; Dumesnil, J.; Veltkamp, A.C.; Grabon, V.; Beech, D.; Buckthorpe, D.; Dominguez, T.; Krussenberg, A.K.; Wu, C.H.

    2000-01-01

    Following an overview on kinetics of carbon/gas reactions, status and further needs in selected safety relevant fields of graphite oxidation in high temperature reactors (HTRs) and fusion reactors are outlined. Kinetics was detected due to the presence of such elements as severe air ingress, lack of experimental data on Boudouard reaction and a similar lack of data in the field of advanced oxidation. The development of coatings which protect against oxidation should focus on stability under neutron irradiation and on the general feasibility of coatings on HTR pebble fuel graphite. Oxidation under normal operation of direct cycle HTR requires examinations of gas atmospheres and of catalytic effects. Advanced carbon materials like CFCs and mixed materials should be developed and tested with respect to their oxidation resistance in a common HTR/fusion task. In an interim HTR, fuel storage radiolytic oxidation under normal operation and thermal oxidation in accidents have to be considered. Plans for future work in these fields are described. (authors)

  10. Status of reduced enrichment programs for research reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  11. Status of reduced enrichment programs for research reactors in Japan

    International Nuclear Information System (INIS)

    Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo

    1997-01-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE

  12. GENP-2, Program System for Integral Reactor Perturbation

    International Nuclear Information System (INIS)

    Boioli, A.; Cecchini, G.P.

    1975-01-01

    1 - Description of problem or function: GENP-2 is a system of programs that use 'generalized perturbation theory' to calculate the perturbations of reactor integral characteristics which can be expressed by means of ratios between linear or bilinear functionals of the real and/or adjoint fluxes (e.g. reaction rate ratios), due to cross section perturbations. 2 - Method of solution: GENP-2 consists of the following codes: DDV, SORCI, CIAP-PMN and GLOBP-2D. DDV calculates the real or adjoint fluxes and power distribution using multigroup diffusion theory in 2-dimensions. SORCI uses the fluxes from DDV to calculate the real and/or adjoint general perturbation sources. CIAP-PMN reads the sources from SORCI and uses them in the real or adjoint generalised importance calculations (2 dimensions, multi- group diffusion). GLOBP-2D uses the importance calculated by CIAP-PMN, and the fluxes calculated by DDV, in generalised perturbation expressions to calculate the perturbation in the quantity of interest. 3 - Restrictions on the complexity of the problem: DDV although variably dimensioned has the following restrictions: - max. number of mesh points 6400; - max. number of mesh points in 1-dimension 81; - max. number of regions 6400; - max. number of energy groups 100; - if power distribution calculated, product of number of groups and number of regions 2500. The other programs have the same restrictions if applicable

  13. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Science.gov (United States)

    Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

    2008-08-01

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  14. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2008-08-31

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  15. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-15

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology.

  16. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-01

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology

  17. Fast reactor development program in France in 1997

    International Nuclear Information System (INIS)

    Langrand, J.C.; Hubert, G.; Marmonier, P.; Del Negro, R.; Saint-Arroman, G.; Coulon, P.; Mesnage, B.; Pillon, S.; Lefevre, J.C.

    1998-01-01

    , in spite of the availability of one and half core. This decision leads to drop the >. The Capra program was reoriented. Now, emphasis is put on burning the minor actinides (MA) and long-lived fusion products (LLFP), which answers the French 1991 law on radioactive wastes; as a consequence, the experimental programme proposed for Phenix is mainly devoted to this objective. This report presents also the status of Rapsodie, and the main achievements on the R and D programme on fast reactors. (author)

  18. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  19. Bioaugmentation of Syntrophic Acetate-Oxidizing Culture in Biogas Reactors Exposed to Increasing Levels of Ammonia

    Science.gov (United States)

    Westerholm, Maria; Levén, Lotta

    2012-01-01

    The importance of syntrophic acetate oxidation for process stability in methanogenic systems operating at high ammonia concentrations has previously been emphasized. In this study we investigated bioaugmentation of syntrophic acetate-oxidizing (SAO) cultures as a possible method for decreasing the adaptation period of biogas reactors operating at gradually increased ammonia concentrations (1.5 to 11 g NH4+-N/liter). Whole stillage and cattle manure were codigested semicontinuously for about 460 days in four mesophilic anaerobic laboratory-scale reactors, and a fixed volume of SAO culture was added daily to two of the reactors. Reactor performance was evaluated in terms of biogas productivity, methane content, pH, alkalinity, and volatile fatty acid (VFA) content. The decomposition pathway of acetate was analyzed by isotopic tracer experiments, and population dynamics were monitored by quantitative PCR analyses. A shift in dominance from aceticlastic methanogenesis to SAO occurred simultaneously in all reactors, indicating no influence by bioaugmentation on the prevailing pathway. Higher abundances of Clostridium ultunense and Tepidanaerobacter acetatoxydans were associated with bioaugmentation, but no influence on Syntrophaceticus schinkii or the methanogenic population was distinguished. Overloading or accumulation of VFA did not cause notable dynamic effects on the population. Instead, the ammonia concentration had a substantial impact on the abundance level of the microorganisms surveyed. The addition of SAO culture did not affect process performance or stability against ammonia inhibition, and all four reactors deteriorated at high ammonia concentrations. Consequently, these findings further demonstrate the strong influence of ammonia on the methane-producing consortia and on the representative methanization pathway in mesophilic biogas reactors. PMID:22923397

  20. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    International Nuclear Information System (INIS)

    Higinbotham, W.A.

    1994-01-01

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 ± 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF

  1. Continuous Photo-Oxidation in a Vortex Reactor: Efficient Operations Using Air Drawn from the Laboratory.

    Science.gov (United States)

    Lee, Darren S; Amara, Zacharias; Clark, Charlotte A; Xu, Zeyuan; Kakimpa, Bruce; Morvan, Herve P; Pickering, Stephen J; Poliakoff, Martyn; George, Michael W

    2017-07-21

    We report the construction and use of a vortex reactor which uses a rapidly rotating cylinder to generate Taylor vortices for continuous flow thermal and photochemical reactions. The reactor is designed to operate under conditions required for vortex generation. The flow pattern of the vortices has been represented using computational fluid dynamics, and the presence of the vortices can be easily visualized by observing streams of bubbles within the reactor. This approach presents certain advantages for reactions with added gases. For reactions with oxygen, the reactor offers an alternative to traditional setups as it efficiently draws in air from the lab without the need specifically to pressurize with oxygen. The rapid mixing generated by the vortices enables rapid mass transfer between the gas and the liquid phases allowing for a high efficiency dissolution of gases. The reactor has been applied to several photochemical reactions involving singlet oxygen ( 1 O 2 ) including the photo-oxidations of α-terpinene and furfuryl alcohol and the photodeborylation of phenyl boronic acid. The rotation speed of the cylinder proved to be key for reaction efficiency, and in the operation we found that the uptake of air was highest at 4000 rpm. The reactor has also been successfully applied to the synthesis of artemisinin, a potent antimalarial compound; and this three-step synthesis involving a Schenk-ene reaction with 1 O 2 , Hock cleavage with H + , and an oxidative cyclization cascade with triplet oxygen ( 3 O 2 ), from dihydroartemisinic acid was carried out as a single process in the vortex reactor.

  2. Development of ODS (oxide dispersion strengthened) ferritic-martensitic steels for fast reactor fuel cladding

    International Nuclear Information System (INIS)

    Ukai, Shigeharu

    2000-01-01

    In order to attain higher burnup and higher coolant outlet temperature in fast reactor, oxide dispersion strengthened (ODS) ferritic-martensitic steels were developed as a long life fuel cladding. The improvement in formability and ductility, which are indispensable in the cold-rolling method for manufacturing cladding tube, were achieved by controlling the microstructure using techniques such as recrystallization heat-treatment and α to γ phase transformation. The ODS ferritic-martensitic cladding tubes manufactured using these techniques have the highest internal creep rupture strength in the world as ferritic stainless steels. Strength level approaches adequate value at 700degC, which meets the requirement for commercial fast reactors. (author)

  3. Overview of welding of oxide dispersion strengthened (ODS) alloys for advanced nuclear reactor applications

    International Nuclear Information System (INIS)

    Kalvala, Prasad Rao; Raja, K.S.; Misra, Manoranjan; Tache, Ricard A.

    2009-01-01

    Oxide dispersion strengthened (ODS) alloys are very promising materials for Generation IV reactors with a potential to be used at elevated temperatures under severe neutron exposure environment. Welding of the ODS alloys is an understudied problem. In this paper, an overview of welding of the ODS alloys useful for advanced nuclear reactor applications is presented. The microstructural changes and the resultant mechanical properties obtained by various solid state welding processes are reviewed. Based on our results on PM2000, an approach for future work on welding of the ODS alloys is suggested. (author)

  4. Optical modeling of nickel-base alloys oxidized in pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Clair, A. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS, Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, 21078 Dijon cedex (France); Foucault, M.; Calonne, O. [Areva ANP, Centre Technique Departement Corrosion-Chimie, 30 Bd de l' industrie, BP 181, 71205 Le Creusot (France); Finot, E., E-mail: Eric.Finot@u-bourgogne.fr [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS, Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, 21078 Dijon cedex (France)

    2012-10-01

    The knowledge of the aging process involved in the primary water of pressurized water reactor entails investigating a mixed growth mechanism in the corrosion of nickel-base alloys. A mixed growth induces an anionic inner oxide and a cationic diffusion parallel to a dissolution-precipitation process forms the outer zone. The in situ monitoring of the oxidation kinetics requires the modeling of the oxide layer stratification with the full knowledge of the optical constants related to each component. Here, we report the dielectric constants of the alloys 600 and 690 measured by spectroscopic ellipsometry and fitted to a Drude-Lorentz model. A robust optical stratification model was determined using focused ion beam cross-section of thin foils examined by transmission electron microscopy. Dielectric constants of the inner oxide layer depleted in chromium were assimilated to those of the nickel thin film. The optical constants of both the spinels and extern layer were determined. - Highlights: Black-Right-Pointing-Pointer Spectroscopic ellipsometry of Ni-base alloy oxidation in pressurized water reactor Black-Right-Pointing-Pointer Measurements of the dielectric constants of the alloys Black-Right-Pointing-Pointer Optical simulation of the mixed oxidation process using a three stack model Black-Right-Pointing-Pointer Scattered crystallites cationic outer layer; linear Ni-gradient bottom layer Black-Right-Pointing-Pointer Determination of the refractive index of the spinel and the Cr{sub 2}O{sub 3} layers.

  5. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  6. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    International Nuclear Information System (INIS)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-01-01

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at ∼95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite (U(IV)) and metaschoepite (U(VI)). This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water (∼95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO 2 and eliminate H 2 generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  7. Secondary organic aerosol from VOC mixtures in an oxidation flow reactor

    Science.gov (United States)

    Ahlberg, Erik; Falk, John; Eriksson, Axel; Holst, Thomas; Brune, William H.; Kristensson, Adam; Roldin, Pontus; Svenningsson, Birgitta

    2017-07-01

    The atmospheric organic aerosol is a tremendously complex system in terms of chemical content. Models generally treat the mixtures as ideal, something which has been questioned owing to model-measurement discrepancies. We used an oxidation flow reactor to produce secondary organic aerosol (SOA) mixtures containing oxidation products of biogenic (α-pinene, myrcene and isoprene) and anthropogenic (m-xylene) volatile organic compounds (VOCs). The resulting volume concentration and chemical composition was measured using a scanning mobility particle sizer (SMPS) and a high-resolution time-of-flight aerosol mass spectrometer (HR-ToF-AMS), respectively. The SOA mass yield of the mixtures was compared to a partitioning model constructed from single VOC experiments. The single VOC SOA mass yields with no wall-loss correction applied are comparable to previous experiments. In the mixtures containing myrcene a higher yield than expected was produced. We attribute this to an increased condensation sink, arising from myrcene producing a significantly higher number of nucleation particles compared to the other precursors. Isoprene did not produce much mass in single VOC experiments but contributed to the mass of the mixtures. The effect of high concentrations of isoprene on the OH exposure was found to be small, even at OH reactivities that previously have been reported to significantly suppress OH exposures in oxidation flow reactors. Furthermore, isoprene shifted the particle size distribution of mixtures towards larger sizes, which could be due to a change in oxidant dynamics inside the reactor.

  8. TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

    1983-01-01

    TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

  9. US/DOE Man-Machine Integration program for liquid metal reactors

    International Nuclear Information System (INIS)

    D'Zmura, A.P.; Seeman, S.E.

    1985-03-01

    The United States Department of Energy (DOE) Man-Machine Integration program was started in 1980 as an addition to the existing Liquid Metal Fast Breeder Reactor safety base technology program. The overall goal of the DOE program is to enhance the operational safety of liquid metal reactors by optimum integration of humans and machines in the overall reactor plant system and by application of the principles of human-factors engineering to the design of equipment, subsystems, facilities, operational aids, procedures and environments. In the four years since its inception the program has concentrated on understanding the control process for Liquid Metal Reactors (LMRs) and on applying advanced computer concepts to this process. This paper describes the products that have been developed in this program, present computer-related programs, and plans for the future

  10. Evaluation of biochars by temperature programmed oxidation/mass spectrometry

    Science.gov (United States)

    Michael Jackson; Thomas Eberhardt; Akwasi Boateng; Charles Mullen; Les Groom

    2013-01-01

    Biochars produced from thermochemical conversions of biomass were evaluated by temperature programmed oxidation (TPO). This technique, used to characterize carbon deposits on petroleum cracking catalysts, provides information on the oxidative stability of carbonaceous solids, where higher temperature reactivity indicates greater structural order, an important property...

  11. Reactor Materials Program probability of indirectly--induced failure of L and P reactor process water piping

    International Nuclear Information System (INIS)

    Daugherty, W.L.

    1988-01-01

    The design basis accident for the Savannah River Production Reactors is the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping material. The Reactor Materials Program was initiated to provide the technical basis for an alternate credible design basis accident. One aspect of this work is to determine the probability of the DEGB; to show that in addition to being incredible, it is also highly improbable. The probability of a DEGB is broken into two parts: failure by direct means, and indirectly-induced failure. Failure of the piping by direct means can only be postulated to occur if an undetected crack grows to the point of instability, causing a large pipe break. While this accident is not as severe as a DEGB, it provides a conservative upper bound on the probability of a direct DEGB of the piping. The second part of this evaluation calculates the probability of piping failure by indirect causes. Indirect failure of the piping can be triggered by an earthquake which causes other reactor components or the reactor building to fall on the piping or pull it from its supports. Since indirectly-induced failure of the piping will not always produce consequences as severe as a DEGB, this gives a conservative estimate of the probability of an indirectly- induced DEGB. This second part, indirectly-induced pipe failure, is the subject of this report. Failure by seismic loads in the piping itself will be covered in a separate report on failure by direct causes. This report provides a detailed evaluation of L reactor. A walkdown of P reactor and an analysis of the P reactor building provide the basis for extending the L reactor results to P reactor

  12. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-01-01

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  13. Education program at the Massachusetts Institute of Technology research reactor for pre-college science teachers

    International Nuclear Information System (INIS)

    Hopkins, G.R.; Fecych, W.; Harling, O.K.

    1989-01-01

    A Pre-College Science Teacher (PCST) Seminar program has been in place at the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory for 4 yr. The purpose of the PCST program is to educate teachers in nuclear technology and to show teachers, and through them the community, the types of activities performed at research reactors. This paper describes the background, content, and results of the MIT PCST program

  14. Air trichloroethylene oxidation in a corona plasma-catalytic reactor

    International Nuclear Information System (INIS)

    Masoomi-Godarzi, S.; Ranji-Burachaloo, H.; Khodadadi, A.A.; Vesali-Naseh, M.; Mortazavi, Y.

    2014-01-01

    The oxidative decomposition of trichloroethylene (TCE; 300 ppm) by non-thermal corona plasma was investigated in dry air at atmospheric pressure and room temperature, both in the absence and presence of catalysts including MnO x , CoO x . The catalysts were synthesized by a co-precipitation method. The morphology and structure of the catalysts were characterized by BET surface area measurement and Fourier Transform Infrared (FTIR) methods. Decomposition of TCE and distribution of products were evaluated by a gas chromatograph (GC) and an FTIR. In the absence of the catalyst, TCE removal is increased with increases in the applied voltage and current intensity. Higher TCE removal and CO 2 selectivity is observed in presence of the corona and catalysts, as compared to those with the plasma alone. The results show that MnO x and CoO x catalysts can dissociate the in-plasma produced ozone to oxygen radicals, which enhances the TCE decomposition. (author)

  15. Air trichloroethylene oxidation in a corona plasma-catalytic reactor

    Science.gov (United States)

    Masoomi-Godarzi, S.; Ranji-Burachaloo, H.; Khodadadi, A. A.; Vesali-Naseh, M.; Mortazavi, Y.

    2014-08-01

    The oxidative decomposition of trichloroethylene (TCE; 300 ppm) by non-thermal corona plasma was investigated in dry air at atmospheric pressure and room temperature, both in the absence and presence of catalysts including MnOx, CoOx. The catalysts were synthesized by a co-precipitation method. The morphology and structure of the catalysts were characterized by BET surface area measurement and Fourier Transform Infrared (FTIR) methods. Decomposition of TCE and distribution of products were evaluated by a gas chromatograph (GC) and an FTIR. In the absence of the catalyst, TCE removal is increased with increases in the applied voltage and current intensity. Higher TCE removal and CO2 selectivity is observed in presence of the corona and catalysts, as compared to those with the plasma alone. The results show that MnOx and CoOx catalysts can dissociate the in-plasma produced ozone to oxygen radicals, which enhances the TCE decomposition.

  16. Calcium oxide/carbon dioxide reactivity in a packed bed reactor of a chemical heat pump for high-temperature gas reactors

    International Nuclear Information System (INIS)

    Kato, Yukitaka; Yamada, Mitsuteru; Kanie, Toshihiro; Yoshizawa, Yoshio

    2001-01-01

    The thermal performance of a chemical heat pump that uses a calcium oxide/carbon dioxide reaction system was discussed as a heat storage system for utilizing heat output from high temperature gas reactors (HTGR). Calcium oxide/carbon dioxide reactivity for the heat pump was measured using a packed bed reactor containing 1.0 kg of reactant. The reactor was capable of storing heat at 900 deg. C by decarbonation of calcium carbonate and generating up to 997 deg. C by carbonation of calcium oxide. The amount of stored heat in the reactor was 800-900 kJ kg -1 . The output temperature of the reactor could be controlled by regulating the carbonation pressure. The thermal storage performance of the reactor was superior to that of conventional sensible heat storage systems. A heat pump using this CaO/CO 2 reactor is expected to contribute to thermal load leveling and to realize highly efficient utilization of HTGR output due to the high heat storage density and high-quality temperature output of the heat pump

  17. Participation in the U.S. Department of Energy Reactor Sharing Program. Progress report

    International Nuclear Information System (INIS)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here

  18. PODESY program for flux mapping of CNA II reactor:

    International Nuclear Information System (INIS)

    Ribeiro Guevara, Sergio

    1988-01-01

    The PODESY program, developed by KWU, calculates the spatial flux distribution of CNA II reactor through a three-dimensional expansion of 90 incore detector measurements. The calculation is made in three steps: a) short-term calculation which considers the control rod positions and it has to be done each time the flux mapping is calculated; b) medium-term calculation which includes local burn-up dependent calculation made by diffusion methods in macro-cell configurations (seven channels in hexagonal distribution), and c) long-term calculation, or macroscopic flux determination, that is a fitting and expansion of measured fluxes, previously corrected by local effects, using the eigen functions of the modified diffusion equation. The paper outlines development of step (c) of the calculation. The incore detectors have been located in the central zone of the core. In order to obtain low errors in the expansion procedure it is necessary to include additional points, whose flux values are assumed to be equivalent to detector measurements. These flux values are calculated with detector measurements and a spatial flux distribution calculated by a PUMA code. This PUMA calculation employs a smooth burn-up distribution (local burn-up variations are considered in step (b) of the whole calculation) representing the state of core evolution at the calculation time. The core evolution referred to ends when the equilibrium core condition is reached. Additionally, a calculation method to be employed in the plant in case of incore detector failures, is proposed. (Author) [es

  19. Design reliability assurance program for Korean next generation reactor

    International Nuclear Information System (INIS)

    Lee, Beom-Su; Han, Jin-Kyu; Na, Jang Hwan; Yoo, Kyung Yeong

    1997-01-01

    The Korean Next Generation Reactor (KNGR) project is to develop standardized nuclear power plant design for the construction of future nuclear power plants in Korea. The main purpose of the KNGR project is to develop the advanced nuclear power plants, which enhance safety and economics significantly through the incorporation of design concepts for severe accident prevention and mitigation, supplementary passive safety concept, simplification and application of modularization and so on. For those, Probabilistic Safety Assessment (PSA) and availability study will be performed at the early stage of the design, and the Design Reliability Assurance Program (D-RAP) is applied in the development of the KNGR to ensure that the safety and availability evaluated in the PSA and availability study at the early phase of the design is maintained through the detailed design, construction, procurement and operation of the plants. This paper presents the D-RAP concept that could be applied at the stage of the basic design of the nuclear power plants, based on the models for the reference plants and/or similar plants. 4 refs., 1 fig

  20. Sulfidogenic biotreatment of synthetic acid mine drainage and sulfide oxidation in anaerobic baffled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bekmezci, Ozan K.; Ucar, Deniz [Harran University, Environmental Engineering Department, Osmanbey Campus, 63000 Sanliurfa (Turkey); Kaksonen, Anna H. [CSIRO Land and Water, Underwood Avenue, Floreat, WA 6014 (Australia); Sahinkaya, Erkan, E-mail: erkansahinkaya@yahoo.com [Harran University, Environmental Engineering Department, Osmanbey Campus, 63000 Sanliurfa (Turkey)

    2011-05-30

    The treatment of synthetic acid mine drainage (AMD) water (pH 3.0-6.5) containing sulfate (3.0-3.5 g L{sup -1}) and various metals (Co, Cu, Fe, Mn, Ni, and Zn) was studied in an ethanol-fed sulfate-reducing 4-compartment anaerobic baffled reactor (ABR) at 32 {sup o}C. The reactor was operated for 160 days at different chemical oxygen demand (COD)/sulfate ratios, hydraulic retention times (HRT), pH, and metal concentrations to study the robustness of the process. The last compartment of the reactor was aerated at different rates to study the bio-oxidation of sulfide to elemental sulfur. The highest sulfate reduction efficiency (88%) was obtained with a feed sulfate concentration of 3.5 g L{sup -1}, COD/sulfate mass ratio of 0.737, feed pH of 3.0 and HRT of 2 days without aeration in the 4th compartment. The corresponding COD removal efficiency was about 92%. The alkalinity produced in the sulfidogenic ethanol oxidation neutralized the acidic mine water from pH 3.0-4.5 to pH 7.0-8.0. Effluent soluble and total heavy metal concentrations were substantially reduced with removal efficiencies generally higher than 99%, except for Mn (25-77%). Limited aeration in the 4th compartment of ABR promoted incomplete oxidation of sulfide to elemental sulfur rather than complete oxidation to sulfate. Depending on the aeration rate and HRT, 32-74% of produced sulfide was oxidized to elemental sulfur. This study demonstrates that by optimizing operating conditions, sulfate reduction, metal removal, alkalinity generation, and excess sulfide oxidation can be achieved in a single ABR treating AMD.

  1. Dynamic\tmodelling of catalytic three-phase reactors for hydrogenation and oxidation processes

    Directory of Open Access Journals (Sweden)

    Salmi T.

    2000-01-01

    Full Text Available The dynamic modelling principles for typical catalytic three-phase reactors, batch autoclaves and fixed (trickle beds were described. The models consist of balance equations for the catalyst particles as well as for the bulk phases of gas and liquid. Rate equations, transport models and mass balances were coupled to generalized heterogeneous models which were solved with respect to time and space with algorithms suitable for stiff differential equations. The aspects of numerical solution strategies were discussed and the procedure was illustrated with three case studies: hydrogenation of aromatics, hydrogenation of aldehydes and oxidation of ferrosulphate. The case studies revealed the importance of mass transfer resistance inside the catalyst pallets as well as the dynamics of the different phases being present in the reactor. Reliable three-phase reactor simulation and scale-up should be based on dynamic heterogeneous models.

  2. Experimental and Kinetic Modeling Study of Ethyl Levulinate Oxidation in a Jet-Stirred Reactor

    KAUST Repository

    Wang, Jui-Yang

    2017-06-01

    A jet-stirred reactor was designed and constructed in the Clean Combustion Research Center (CCRC) at King Abdullah University of Science and Technology (KAUST); was validated with n-heptane, iso-octane oxidation and cyclohexene pyrolysis. Different configurations of the setup have been tested to achieve good agreement with results from the literature. Test results of the reactor indicated that installation of a pumping system at the downstream side in the experimental apparatus was necessary to avoid the reoccurrence of reactions in the sampling probe. Experiments in ethyl levulinate oxidation were conducted in the reactor under several equivalence ratios, from 600 to 1000 K, 1 bar and 2 s residence time. Oxygenated species detected included methyl vinyl ketone, levulinic acid and ethyl acrylate. Ethylene, methane, carbon monoxide, hydrogen, oxygen and carbon dioxide were further quantified with a gas chromatography, coupled with a flame ionization detector and a thermal conductivity detector. The ethyl levulinate chemical kinetic model was first developed by Dr. Stephen Dooley, Trinity College Dublin, and simulated under the same conditions, using the Perfect-Stirred Reactor code in Chemkin software. In comparing the simulation results with experimental data, some discrepancies were noted; predictions of ethylene production were not well matched. The kinetic model was improved by updating several classes of reactions: unimolecular decomposition, H-abstraction, C-C and C-O beta-scissions of fuel radicals. The updated model was then compared again with experimental results and good agreement was achieved, proving that the concerted eliminated reaction is crucial for the kinetic mechanism formulation of ethyl levulinate. In addition, primary reaction pathways and sensitivity analysis were performed to describe the role of molecular structure in combustion (800 and 1000 K for ethyl levulinate oxidation in the jet-stirred reactor).

  3. Progress of the United States foreign research reactor spent nuclear fuel acceptance program. Reduced enrichment for research and test reactors conference 2002

    International Nuclear Information System (INIS)

    Clapper, Maureen

    2002-01-01

    Foreign Research Reactor Spent nuclear fuel Acceptance Program is actively working with research reactors to accept eligible material before the Acceptance Policy proper expires in 2006. Reactors/governments wishing to participate should contact US immediately if they have not done so already. Program operations are changing to adapt to new challenges. We continue to promote the importance of this Program to senior management in the Department of Energy

  4. The U.S. DOE new production reactor/heavy water reactor facility pollution prevention/waste minimization program

    International Nuclear Information System (INIS)

    Kaczmarsky, Myron M.; Tsang, Irving; Stepien, Walter P.

    1992-01-01

    A Pollution Prevention/Waste Minimization Program was established during the early design phase of the U.S. DOE's New Production Reactor/Heavy Water Reactor Facility (NPR/HWRF) to encompass design, construction, operation and decommissioning. The primary emphasis of the program was given to waste elimination, source reduction and/or recycling to minimize the quantity and toxicity of material before it enters the waste stream for treatment or disposal. The paper discusses the regulatory and programmatic background as it applies to the NPR/HWRF and the waste assessment program developed as a phased approach to pollution prevention/waste minimization for the NPR/HWRF. Implementation of the program will be based on various factors including life cycle cost analysis, which will include costs associated with personnel, record keeping, transportation, pollution control equipment, treatment, storage, disposal, liability, compliance and oversight. (author)

  5. Monte Carlo radiative transfer simulation of a cavity solar reactor for the reduction of cerium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Villafan-Vidales, H.I.; Arancibia-Bulnes, C.A.; Dehesa-Carrasco, U. [Centro de Investigacion en Energia, Universidad Nacional Autonoma de Mexico, Privada Xochicalco s/n, Col. Centro, A.P. 34, Temixco, Morelos 62580 (Mexico); Romero-Paredes, H. [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco No.186, Col. Vicentina, A.P. 55-534, Mexico D.F 09340 (Mexico)

    2009-01-15

    Radiative heat transfer in a solar thermochemical reactor for the thermal reduction of cerium oxide is simulated with the Monte Carlo method. The directional characteristics and the power distribution of the concentrated solar radiation that enters the cavity is obtained by carrying out a Monte Carlo ray tracing of a paraboloidal concentrator. It is considered that the reactor contains a gas/particle suspension directly exposed to concentrated solar radiation. The suspension is treated as a non-isothermal, non-gray, absorbing, emitting, and anisotropically scattering medium. The transport coefficients of the particles are obtained from Mie-scattering theory by using the optical properties of cerium oxide. From the simulations, the aperture radius and the particle concentration were optimized to match the characteristics of the considered concentrator. (author)

  6. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Cahalan, J.; Wigeland, R.; Friedel, G.; Kussmaul, G.; Royl, P.; Moreau, J.; Perks, M.

    1990-01-01

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs

  7. [Kinetics of catalytic wet air oxidation of phenol in trickle bed reactor].

    Science.gov (United States)

    Li, Guang-ming; Zhao, Jian-fu; Wang, Hua; Zhao, Xiu-hua; Zhou, Yang-yuan

    2004-05-01

    By using a trickle bed reactor which was designed by the authors, the catalytic wet air oxidation reaction of phenol on CuO/gamma-Al2O3 catalyst was studied. The results showed that in mild operation conditions (at temperature of 180 degrees C, pressure of 3 MPa, liquid feed rate of 1.668 L x h(-1) and oxygen feed rate of 160 L x h(-1)), the removal of phenol can be over 90%. The curve of phenol conversion is similar to "S" like autocatalytic reaction, and is accordance with chain reaction of free radical. The kinetic model of pseudo homogenous reactor fits the catalytic wet air oxidation reaction of phenol. The effects of initial concentration of phenol, liquid feed rate and temperature for reaction also were investigated.

  8. On-chip microplasma reactors using carbon nanofibres and tungsten oxide nanowires as electrodes

    International Nuclear Information System (INIS)

    Agiral, Anil; Groenland, Alfons W; Han Gardeniers, J G E; Chinthaginjala, J Kumar; Seshan, K; Lefferts, Leon

    2008-01-01

    Carbon nanofibres (CNFs) and tungsten oxide (W 18 O 49 ) nanowires have been incorporated into a continuous flow type microplasma reactor to increase the reactivity and efficiency of the barrier discharge at atmospheric pressure. CNFs and tungsten oxide nanowires were characterized by high-resolution scanning electron microscopy, transmission electron microscopy and nanodiffraction methods. Field emission of electrons from those nanostructures supplies free electrons and ions during microplasma production. Reduction in breakdown voltage, higher number of microdischarges and higher energy deposition were observed at the same applied voltage when compared with plane electrodes at atmospheric pressure in air. Rate coefficients of electron impact reaction channels to decompose CO 2 were calculated and it was shown that CO 2 consumption increased using CNFs compared with plane electrode in the microplasma reactor.

  9. Final report. U.S. Department of Energy University Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, John A

    2003-01-21

    Activities supported at the MIT Nuclear Reactor Laboratory under the U.S. DOE University Reactor Sharing Program are reported for Grant DE FG02-95NE38121 (September 16, 1995 through May 31, 2002). These activities fell under four subcategories: support for research at thesis and post-doctoral levels, support for college-level laboratory exercises, support for reactor tours/lectures on nuclear energy, and support for science fair participants.

  10. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor

    Energy Technology Data Exchange (ETDEWEB)

    Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: vaveri@kist.re.kr; Kim, Jae-Duck [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: jdkim@kist.re.kr; Lee, Jong-Chol [Agency for Defense Development (ADD), P.O. Box 35-1, Yuseong-gu, Daejeon (Korea, Republic of)]. E-mail: jcleeadd@hanafos.com

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG (HOC{sub 2}H{sub 4}){sub 2}S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

  11. Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

    2014-01-01

    Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

  12. Trickle bed reactor for the oxidation of phenol over active carbon catalyst

    OpenAIRE

    Gabbiye, Nigus; Font Capafons, Josep; Fortuny Sanromá, Agustín; Bengoa, Christophe José; Fabregat Llangotera, Azael; Stüber, Frank Erich

    2009-01-01

    The catalytic wet air oxidation of phenol using activated carbon has been performed in a laboratory trickle bed reactor over a wide range of operating variables (PO2, T, FL and Cph,o) and hydrodynamic conditions. The influence of different start-up procedures (saturation of activated carbon) has also been tested. Further improvement of activity and stability has been checked for by using dynamic TBR operation concept or impregnated Fe/carbon catalyst. The results obtained confi...

  13. Radiological control aspects of the fabrication of the Light Water Breeder Reactor core (LWBR Development Program)

    International Nuclear Information System (INIS)

    Schultz, B.G.

    1979-05-01

    A description is presented of the radiological control aspects of the fabrication of the Light Water Breeder Reactor (LWBR) core. Included are the radiological control criteria applied for the design and use of fabrication facilities, the controls and limits imposed to minimize radiaion exposure to personnel, and an evaluation of the applied radiological program in meeting the program objectives. The goal of the LWBR program is to develop the technology to breed in light water reactors so that nuclear fuel may be used significantly more efficiently in these reactors. This technology is being developed by designing and fabricating a breeder reactor core, utilizing thoria (ThO 2 ) and binary thoria--urania (ThO 2 - 233 UO 2 ) fuel, to be operated in the existing pressurized water reactor plant owned by the Department of Energy at Shippingport, Pennsylvania

  14. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

    1980-03-01

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  15. Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.; Johnson, Gerald D.; Brown, W. F.; Paxton, Michael M.; Puigh, Raymond J.; Eiholzer, Cheryl R.; Martinez, C.; Blotter, M. A.

    2000-02-28

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

  16. Development of a membrane-assisted fluidized bed reactor - 2 - Experimental demonstration and modeling for the partial oxidation of methanol

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.; Laverman, J.A.; van Sint Annaland, M.; Kuipers, J.A.M.

    2005-01-01

    A small laboratory-scale membrane-assisted fluidized bed reactor (MAFBR) was constructed in order to experimentally demonstrate the reactor concept for the partial oxidation of methanol to formaldehyde. Methanol conversion and product selectivities were measured at various overall fluidization

  17. The gas-solid trickle-flow reactor for the catalytic oxidation of hydrogen sulphide: a trickle-phase model

    NARCIS (Netherlands)

    Verver, A.B.; van Swaaij, Willibrordus Petrus Maria

    1987-01-01

    The oxidation of H2S by O2 producing elemental sulphur has been studied at temperatures of 100–300°C and at atmospheric pressure in a laboratory-scale gas-solid trickle-flow reactor. In this reactor one of the reaction products, i.e. sulphur, is removed continuously by flowing solids. A porous,

  18. Investigation on the combined operation of water gas shift and preferential oxidation reactor system on the kW scale

    NARCIS (Netherlands)

    O'Connell, M.; Kolb, G.A.; Schelhaas, K.P.; Schuerer, J.; Tiemann, D.; Keller, S.; Reinhard, D.; Hessel, V.

    2010-01-01

    A 5 kWel water gas shift reactor was integrated with a 5 kWel preferential oxidation reactor for the purposes of reducing the carbon monoxide levels in a reformate exit stream to levels below 100 ppm. The integrated system worked best at partial load with CO concentrations being reduced to 40 ppm at

  19. University Reactor Sharing Program. Final report, September 30, 1992--September 29, 1994

    International Nuclear Information System (INIS)

    Wehring, B.W.

    1995-01-01

    Over the past 20 years, the number of nuclear reactors on university campuses in the US declined from more than 70 to less than 40. Contrary to this trend, The University of Texas at Austin constructed a new reactor facility at a cost of $5.8 million. The new reactor facility houses a new TRIGA Mark II reactor which replaces an in-ground TRIGA Mark I reactor located in a 50-year old building. The new reactor facility was constructed to strengthen the instruction and research opportunities in nuclear science and engineering for both undergraduate and graduate students at The University of Texas. On January 17, 1992, The University of Texas at Austin received a license for operation of the new reactor. Initial criticality was achieved on March 12, 1992, and full power operation, on March 25, 1992. The UT-TRIGA research reactor provides hands-on education, multidisciplinary research and unique service activities for academic, medical, industrial, and government groups. Support by the University Reactor Sharing Programs increases the availability of The University of Texas reactor facility for use by other educational institutions which do not have nuclear reactors

  20. Fabrication technological development of the oxide dispersion strengthened alloy MA957 for fast reactor applications

    International Nuclear Information System (INIS)

    ML Hamilton; DS Gelles; RJ Lobsinger; GD Johnson; WF Brown; MM Paxton; RJ Puigh; CR Eiholzer; C Martinez; MA Blotter

    2000-01-01

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material, in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report. The alloy is a ferritic stainless steel developed by International Nickel Company specifically for structural reactor applications. It is strengthened by a very fine, uniformly distributed yttria dispersoid. Its fabrication involves a mechanical alloying process and subsequent extrusion, which ultimately results in a highly elongated grain structure. While the presence of the dispersoid produces a material with excellent strength, the body centered cubic structure inherent to the material coupled with the high aspect ratio that results from processing operations produces some difficulties with ductility. The alloy is very sensitive to variations in a number of processing parameters, and if the high strength is once lost during fabrication, it cannot be recovered. The microstructural evolution of the alloy under irradiation falls into two regimes. Below about 550 C, dislocation development, αprime precipitation and void evolution in the matrix are observed, while above about 550 C damage appears to be restricted to cavity formation within oxide particles. The thermal expansion of the alloy is very similar to that of HT9 up to the temperature where HT9 undergoes a phase transition to austenitic. Pulse magnetic welding of end caps onto MA957 tubing can be accomplished in a manner similar to that in which it is performed on HT9, although the welding parameters appear to be very sensitive to variations in the tubing that result from small changes in fabrication conditions. The tensile and stress rupture behavior of the alloy are acceptable in the unirradiated condition, being comparable to HT9 below about 700 C and exceeding those of HT9 at

  1. Fabrication technological development of the oxide dispersion strengthened alloy MA957 for fast reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    ML Hamilton; DS Gelles; RJ Lobsinger; GD Johnson; WF Brown; MM Paxton; RJ Puigh; CR Eiholzer; C Martinez; MA Blotter

    2000-03-27

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material, in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report. The alloy is a ferritic stainless steel developed by International Nickel Company specifically for structural reactor applications. It is strengthened by a very fine, uniformly distributed yttria dispersoid. Its fabrication involves a mechanical alloying process and subsequent extrusion, which ultimately results in a highly elongated grain structure. While the presence of the dispersoid produces a material with excellent strength, the body centered cubic structure inherent to the material coupled with the high aspect ratio that results from processing operations produces some difficulties with ductility. The alloy is very sensitive to variations in a number of processing parameters, and if the high strength is once lost during fabrication, it cannot be recovered. The microstructural evolution of the alloy under irradiation falls into two regimes. Below about 550 C, dislocation development, {alpha}{prime} precipitation and void evolution in the matrix are observed, while above about 550 C damage appears to be restricted to cavity formation within oxide particles. The thermal expansion of the alloy is very similar to that of HT9 up to the temperature where HT9 undergoes a phase transition to austenitic. Pulse magnetic welding of end caps onto MA957 tubing can be accomplished in a manner similar to that in which it is performed on HT9, although the welding parameters appear to be very sensitive to variations in the tubing that result from small changes in fabrication conditions. The tensile and stress rupture behavior of the alloy are acceptable in the unirradiated condition, being comparable to HT9 below about 700 C and exceeding those of HT9

  2. Isotopic evidence for nitrous oxide production pathways in a partial nitritation-anammox reactor.

    Science.gov (United States)

    Harris, Eliza; Joss, Adriano; Emmenegger, Lukas; Kipf, Marco; Wolf, Benjamin; Mohn, Joachim; Wunderlin, Pascal

    2015-10-15

    Nitrous oxide (N2O) production pathways in a single stage, continuously fed partial nitritation-anammox reactor were investigated using online isotopic analysis of offgas N2O with quantum cascade laser absorption spectroscopy (QCLAS). N2O emissions increased when reactor operating conditions were not optimal, for example, high dissolved oxygen concentration. SP measurements indicated that the increase in N2O was due to enhanced nitrifier denitrification, generally related to nitrite build-up in the reactor. The results of this study confirm that process control via online N2O monitoring is an ideal method to detect imbalances in reactor operation and regulate aeration, to ensure optimal reactor conditions and minimise N2O emissions. Under normal operating conditions, the N2O isotopic site preference (SP) was much higher than expected - up to 40‰ - which could not be explained within the current understanding of N2O production pathways. Various targeted experiments were conducted to investigate the characteristics of N2O formation in the reactor. The high SP measurements during both normal operating and experimental conditions could potentially be explained by a number of hypotheses: i) unexpectedly strong heterotrophic N2O reduction, ii) unknown inorganic or anammox-associated N2O production pathway, iii) previous underestimation of SP fractionation during N2O production from NH2OH, or strong variations in SP from this pathway depending on reactor conditions. The second hypothesis - an unknown or incompletely characterised production pathway - was most consistent with results, however the other possibilities cannot be discounted. Further experiments are needed to distinguish between these hypotheses and fully resolve N2O production pathways in PN-anammox systems. Copyright © 2015 Elsevier Ltd. All rights reserved.

  3. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress for the past twelve months and will be continued this fiscal year, from April 1982 through March 1983, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1981. The 1982 year budget for R and D work and for construction of a prototype fast breeder reactor MONJU is approximately 20 and 27 billion yen respectively, excluding wages for the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaged in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and six operational cycles at 75 MWt were completed in December 1981. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor was approved by the authorities concerned, and the licensing of the first step was completed at the end of 1981. Preliminary design studies of a large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  4. Developing a framework for a sustainable research reactor program in the UAE

    Energy Technology Data Exchange (ETDEWEB)

    Almarri, Khalid [Senate Member and Representative of Students of PhD Program in Project Management, Ajman (United Arab Emirates)

    2013-07-01

    In 2009, the UAE passed a milestone of preparations for the involvement in the nuclear area by awarding its first nuclear power plant. And to maximise its nuclear benefits, the country needs to develop a research reactor program. This paper's objectives are: a) Selecting and analysing model case studies from other nations to establish the success factors of research reactors in the UAE, and the preparedness of local industries to maximise the potential of the reactors. b) Establishing how the UAE's research reactors will contribute to the coalition and sharing of good practices with other countries as promoted by IAEA.

  5. Progress and status of the Integral Fast Reactor (IFR) development program

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1992-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this

  6. Progress and status of the Integral Fast Reactor (IFR) development program

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1992-04-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this.

  7. Progress and status of the Integral Fast Reactor (IFR) development program

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1992-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this.

  8. Nuclear power reactor analysis, methods, algorithms and computer programs

    International Nuclear Information System (INIS)

    Matausek, M.V

    1981-01-01

    Full text: For a developing country buying its first nuclear power plants from a foreign supplier, disregarding the type and scope of the contract, there is a certain number of activities which have to be performed by local stuff and domestic organizations. This particularly applies to the choice of the nuclear fuel cycle strategy and the choice of the type and size of the reactors, to bid parameters specification, bid evaluation and final safety analysis report evaluation, as well as to in-core fuel management activities. In the Nuclear Engineering Department of the Boris Kidric Institute of Nuclear Sciences (NET IBK) the continual work is going on, related to the following topics: cross section and resonance integral calculations, spectrum calculations, generation of group constants, lattice and cell problems, criticality and global power distribution search, fuel burnup analysis, in-core fuel management procedures, cost analysis and power plant economics, safety and accident analysis, shielding problems and environmental impact studies, etc. The present paper gives the details of the methods developed and the results achieved, with the particular emphasis on the NET IBK computer program package for the needs of planning, construction and operation of nuclear power plants. The main problems encountered so far were related to small working team, lack of large and powerful computers, absence of reliable basic nuclear data and shortage of experimental and empirical results for testing theoretical models. Some of these difficulties have been overcome thanks to bilateral and multilateral cooperation with developed countries, mostly through IAEA. It is the authors opinion, however, that mutual cooperation of developing countries, having similar problems and similar goals, could lead to significant results. Some activities of this kind are suggested and discussed. (author)

  9. Artificial intelligence program in a computer application supporting reactor operations

    International Nuclear Information System (INIS)

    Stratton, R.C.; Town, G.G.

    1985-01-01

    Improving nuclear reactor power plant operability is an ever-present concern for the nuclear industry. The definition of plant operability involves a complex interaction of the ideas of reliability, safety, and efficiency. This paper presents observations concerning the issues involved and the benefits derived from the implementation of a computer application which combines traditional computer applications with artificial intelligence (AI) methodologies. A system, the Component Configuration Control System (CCCS), is being installed to support nuclear reactor operations at the Experimental Breeder Reactor II

  10. An overview of the Indian program related to fast reactor core mechanical behaviour

    International Nuclear Information System (INIS)

    Govindarajan, S.; Bhoje, S.B.; Paranjpe, S.R.

    1984-01-01

    This Indian review paper presents the evolution of the fast breeder program which began with fast breeder test reactor (FBTR) commencing in 1972. The state-of-art in the field of core mechanical behaviour is reviewed

  11. LTFR-4, Library Generated for Fast Reactor Design Program from JAERI Fast-Set Multigroup Constant

    International Nuclear Information System (INIS)

    Suzuki, Tomoo

    1971-01-01

    Nature of physical problem solved: The program processes JAERI-Fast group constants sets of less than 30-group and prepares a binary library tape for efficient usage by a series of related fast reactor design calculation programmes

  12. Nitrous Oxide Production in a Granule-based Partial Nitritation Reactor: A Model-based Evaluation.

    Science.gov (United States)

    Peng, Lai; Sun, Jing; Liu, Yiwen; Dai, Xiaohu; Ni, Bing-Jie

    2017-04-03

    Sustainable wastewater treatment has been attracting increasing attentions over the past decades. However, the production of nitrous oxide (N 2 O), a potent GHG, from the energy-efficient granule-based autotrophic nitrogen removal is largely unknown. This study applied a previously established N 2 O model, which incorporated two N 2 O production pathways by ammonia-oxidizing bacteria (AOB) (AOB denitrification and the hydroxylamine (NH 2 OH) oxidation). The two-pathway model was used to describe N 2 O production from a granule-based partial nitritation (PN) reactor and provide insights into the N 2 O distribution inside granules. The model was evaluated by comparing simulation results with N 2 O monitoring profiles as well as isotopic measurement data from the PN reactor. The model demonstrated its good predictive ability against N 2 O dynamics and provided useful information about the shift of N 2 O production pathways inside granules for the first time. The simulation results indicated that the increase of oxygen concentration and granule size would significantly enhance N 2 O production. The results further revealed a linear relationship between N 2 O production and ammonia oxidation rate (AOR) (R 2  = 0.99) under the conditions of varying oxygen levels and granule diameters, suggesting that bulk oxygen and granule size may exert an indirect effect on N 2 O production by causing a change in AOR.

  13. Reactor vessel assessment and the development of a reactor vessel life extension program for Calvert Cliffs Units One and Two

    International Nuclear Information System (INIS)

    Montgomery, B.; Hijeck, P.J.

    1988-01-01

    A study has been undertaken to provide a general assessment of the life extension capabilities for the Calvert Cliffs Units One and Two reactor pressure vessels. The purpose of the study is to assess the general life extension capabilities for the Calvert Cliffs reactor pressure vessels based upon an extension and variation of the Surry pilot plant life extension study. This assessment provided a detailed reactor vessel surveillance program for plant life extension along with a hierarchy of specific tasks necessary for attaining maximum useful life. The assessment identified a number of critical issues which may impact life attainment and extension along with potential solutions to address these issues to ensure the life extension option is not precluded

  14. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

  15. Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

    Science.gov (United States)

    Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

    2014-01-01

    The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

  16. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  17. GCRA review and appraisal of HTGR reactor-core-design program

    International Nuclear Information System (INIS)

    1980-09-01

    The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation

  18. Possible effects of UO2 oxidation on light water reactor spent fuel performance in long-term geologic disposal

    International Nuclear Information System (INIS)

    Almassy, M.Y.; Woodley, R.E.

    1982-08-01

    Disposal of spent nuclear fuel in a conventionally mined geologic formation is the nearest-term option for permanently isolating radionuclides from the biosphere. Because irradiated uranium dioxide (UO 2 ) fuel pellets retain 95 to 99% of the radionuclides generated during normal light water reactor operation, they may represent a significant barrier to radionuclide release. This document presents a technical assessment of published literature representing the current level of understanding of spent fuel characteristics and conditions that may degrade pellet integrity during a geologic disposal sequence. A significant deterioration mechanism is spent UO 2 oxidation with possible consequences identified as fission gas release, rod diameter increases, cladding breach extension, and release of solid fuel particles containing radionuclides. Areas requiring further study to support development of a comprehensive spent fuel performance prediction model are highlighted. A program and preliminary schedule to obtain the information needed to develop model correlations are also presented

  19. Quantification of chlorine in zirconium oxide and biological samples by instrumental NAA utilizing PCF of Dhruva reactor

    International Nuclear Information System (INIS)

    Shinde, Amol D.; Reddy, A.V.R.; Acharya, R.; Balaji Rao, Y.

    2012-01-01

    Recently studies on chlorine contents in various samples are being pursued due to its corrosive nature. Chlorine present at trace level in various finished products as well as powder is used as a raw material for production of different types of zircaloys used as structural materials in nuclear technology. As a part of quality assurance program, it is necessary to quantify chlorine accurately with suitable and simple technique. In the present work we have applied instrumental neutron activation analysis (INAA) utilizing its short-lived activation product ( 38 Cl, 37 min, 1642 and 2168 keV) for its estimation. Pneumatic Carrier Facility (PCF) of Dhruva reactor, BARC was used sample irradiation of zirconium oxide dry powder, synthetic wax and IAEA RMs 1515 (Apple leaves) and Lichen 336. (author)

  20. MSR - SPHINX concept program Eros (Experimental zero power Salt reactor SR-0) - The proposed experimental program as a basis for validation of reactor physics methods

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Juricek, V.; Kyncl, J.; Mikisek, M.; Rypar, V. [Nuclear Research Institute Rez plc, Rez (Czech Republic)

    2007-07-01

    The Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept solves this principal problem of spent fuel treatment by means of so-called nuclear incineration. It means the burning of fissionable part of its inventory and transmutation of other problematic radionuclides by use of nuclear reactions with neutrons in a MSR-SPHINX system. This reactor system is an actinide burner (most in resonance neutron spectrum) and a radionuclide transmuter in a well-thermalized neutron spectrum. In the frame of the physical part, there are computational analyses and experimental activities. The experimental program has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors. The proposed next stage of the program will focus on a large-scale experimental verification of design inputs by use of MSR-type inserting zones into the existing light water moderated experimental reactor LR-0, which may allow us to modify it into the experimental zero power salt reactor SR-0. There will be a detail description of the proposed program given in the paper together with the so far performed experiments and their first results. These realized experiments help us also to verify computational codes used, and to recognize some anomalies related to molten fluorides utilization. (authors)

  1. Sodium components cleaning status in the Italian fast reactor program

    Energy Technology Data Exchange (ETDEWEB)

    De Luca, B [CNEN-RIT/MAT - Laboratorio Sviluppo Processi - C.S.N. Cassacia, Rome (Italy); Labanti, V [CNEN-DRV, Bologna (Italy); Mennucci, M [NIRA, Genoa (Italy)

    1978-08-01

    As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed.

  2. Partial oxidation of methane in a temperature-controlled dielectric barrier discharge reactor

    KAUST Repository

    Zhang, Xuming

    2015-01-01

    We studied the relative importance of the reduced field intensity and the background reaction temperature in the partial oxidation of methane in a temperature-controlled dielectric barrier discharge reactor. We obtained important mechanistic insight from studying high-temperature and low-pressure conditions with similar reduced field intensities. In the tested range of background temperatures (297 < T < 773 K), we found that the conversion of methane and oxygen depended on both the electron-induced chemistry and the thermo-chemistry, whereas the chemical pathways to the products were overall controlled by the thermo-chemistry at a given temperature. We also found that the thermo-chemistry enhanced the plasma-assisted partial oxidation process. Our findings expand our understanding of the plasma-assisted partial oxidation process and may be helpful in the design of cost-effective plasma reformers. © 2014 The Combustion Institute.

  3. Data-processing program from the operating modes of the nuclear reactor (P0DER)

    International Nuclear Information System (INIS)

    Totev, T.L.; Boyadzhiev, A.I.

    1981-01-01

    A program PODER for processing data from the operating modes of the reactors taking into account the effects of corrosion, hydration, and deformation of the nuclear reactor fuel element sheathing, the formation of the corrosion product deposits, the change in the geometric dimensions of the nuclear reactor fuel element due to the temperature deformation, as well as the various gas fillers, are elaborated. The ''hot channel'' method determining the reliability of the system is realized. The basic equations describing the thermohydraulic processes in nuclear reactors are solved by the finite difference method. Approximations are carried out with the approach of least squares. The temperature distribution versus the zirconium sheathing height is computed for the case of WWER-440 type reactors. The advantages of the proposed program P0DER are discussed

  4. Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.

    1987-01-01

    The Argentine Atomic Energy Commission started in 1978 the Reduced Enrichment Research and Test Reactors in the field of reactor engineering; engineering, development and manufacturing of fuel elements and research reactors operators. This program was initiated with the conviction that it would contribute to the international efforts to reduce risks of nuclear weapons proliferation owing to an uncontrolled use of highly enriched uranium. It was intended to convert RA-3 reactor to make possible its operation with low enriched fuel (LEU), instead of high enriched fuel (HEU) and to develop manufacturing techniques for said LEU. Afterwards, this program was adapted to assist other countries in reactors conversion, development of the corresponding fuel elements and supply of fuel elements to other countries. (Author)

  5. Test program for NIS calibration to reactor thermal output in HTTR

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Shinozaki, Masayuki; Tachibana, Yukio; Kunitomi, Kazuhiko

    2000-03-01

    Rise-to-power test program for reactor thermal output measurement has been established to calibrate a neutron instrumentation system taking account of the characteristics of the High Temperature Engineering Test Reactor (HTTR). An error of reactor thermal output measurement was evaluated taking account of a configuration of instrumentation system. And the expected dispersion of measurement in the full power operation was evaluated from non-nuclear heat-up of primary coolant up to 213degC. From the evaluation, it was found that an error of reactor thermal output measurement would be less than ±2.0% at the rated power. This report presents the detailed program of rise-to-power test for reactor thermal output measurement and discusses its measurement error. (author)

  6. Application of the REMIX thermal mixing calculation program for the Loviisa reactor

    International Nuclear Information System (INIS)

    Kokkonen, I.; Tuomisto, H.

    1987-08-01

    The REMIX computer program has been validated to be used in the pressurized thermal shock study of the Loviisa reactor pressure vessel. The program has been verified against the data from the thermal and fluid mixing experiments. These experiments have been carried out in Imatran voima Oy to study thermal mixing of the high-pressure safety injection water in the Loviisa VVER-440 type pressurized water reactor. The verified REMIX-versions were applied to reactor calculations in the probabilistic pressurized thermal shock study of the Loviisa Plant

  7. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    OpenAIRE

    KO, DO-YOUNG; KIM, KYU-HYUNG

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

  8. Treatment of textile effluent by chemical (Fenton's Reagent) and biological (sequencing batch reactor) oxidation

    International Nuclear Information System (INIS)

    Rodrigues, Carmen S.D.; Madeira, Luis M.; Boaventura, Rui A.R.

    2009-01-01

    The removal of organic compounds and colour from a synthetic effluent simulating a cotton dyeing wastewater was evaluated by using a combined process of Fenton's Reagent oxidation and biological degradation in a sequencing batch reactor (SBR). The experimental design methodology was first applied to the chemical oxidation process in order to determine the values of temperature, ferrous ion concentration and hydrogen peroxide concentration that maximize dissolved organic carbon (DOC) and colour removals and increase the effluent's biodegradability. Additional studies on the biological oxidation (SBR) of the raw and previously submitted to Fenton's oxidation effluent had been performed during 15 cycles (i.e., up to steady-state conditions), each one with the duration of 11.5 h; Fenton's oxidation was performed either in conditions that maximize the colour removal or the increase in the biodegradability. The obtained results allowed concluding that the combination of the two treatment processes provides much better removals of DOC, BOD 5 and colour than the biological or chemical treatment alone. Moreover, the removal of organic matter in the integrated process is particularly effective when Fenton's pre-oxidation is carried out under conditions that promote the maximum increase in wastewater biodegradability.

  9. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  10. Molecular characterization of anaerobic sulfur-oxidizing microbial communities in up-flow anaerobic sludge blanket reactor treating municipal sewage.

    Science.gov (United States)

    Aida, Azrina A; Hatamoto, Masashi; Yamamoto, Masamitsu; Ono, Shinya; Nakamura, Akinobu; Takahashi, Masanobu; Yamaguchi, Takashi

    2014-11-01

    A novel wastewater treatment system consisting of an up-flow anaerobic sludge blanket (UASB) reactor and a down-flow hanging sponge (DHS) reactor with sulfur-redox reaction was developed for treatment of municipal sewage under low-temperature conditions. In the UASB reactor, a novel phenomenon of anaerobic sulfur oxidation occurred in the absence of oxygen, nitrite and nitrate as electron acceptors. The microorganisms involved in anaerobic sulfur oxidation have not been elucidated. Therefore, in this study, we studied the microbial communities existing in the UASB reactor that probably enhanced anaerobic sulfur oxidation. Sludge samples collected from the UASB reactor before and after sulfur oxidation were used for cloning and terminal restriction fragment length polymorphism (T-RFLP) analysis of the 16S rRNA genes of the bacterial and archaeal domains. The microbial community structures of bacteria and archaea indicated that the genus Smithella and uncultured bacteria within the phylum Caldiserica were the dominant bacteria groups. Methanosaeta spp. was the dominant group of the domain archaea. The T-RFLP analysis, which was consistent with the cloning results, also yielded characteristic fingerprints for bacterial communities, whereas the archaeal community structure yielded stable microbial community. From these results, it can be presumed that these major bacteria groups, genus Smithella and uncultured bacteria within the phylum Caldiserica, probably play an important role in sulfur oxidation in UASB reactors. Copyright © 2014 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  11. Homogeneity of Continuum Model of an Unsteady State Fixed Bed Reactor for Lean CH4 Oxidation

    Directory of Open Access Journals (Sweden)

    Subagjo

    2014-07-01

    Full Text Available In this study, the homogeneity of the continuum model of a fixed bed reactor operated in steady state and unsteady state systems for lean CH4 oxidation is investigated. The steady-state fixed bed reactor system was operated under once-through direction, while the unsteady-state fixed bed reactor system was operated under flow reversal. The governing equations consisting of mass and energy balances were solved using the FlexPDE software package, version 6. The model selection is indispensable for an effective calculation since the simulation of a reverse flow reactor is time-consuming. The homogeneous and heterogeneous models for steady state operation gave similar conversions and temperature profiles, with a deviation of 0.12 to 0.14%. For reverse flow operation, the deviations of the continuum models of thepseudo-homogeneous and heterogeneous models were in the range of 25-65%. It is suggested that pseudo-homogeneous models can be applied to steady state systems, whereas heterogeneous models have to be applied to unsteady state systems.

  12. FFTF reactor-characterization program: gamma-ray measurements and shield characterization

    International Nuclear Information System (INIS)

    Bunch, W.L.; Moore, F.S. Jr.

    1983-02-01

    A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA)

  13. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  14. A review of fast reactor program in Japan - April 1984

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1984-01-01

    The fast breeder reactor development project in PNC has been in progress steadily in these eighteen years. Concerning the experimental fast reactor, JOYO, the MK-II core attained criticality on November 22, 1982 with 51 fuel assemblies, and received the ''Certificate of Inspection before Operation'' from Government Authority on March 31, 1983, after 100 hours operation with the rated output of 100 MW. Since then, the core has been utilized to implement irradiation bed characteristics test, and to irradiate fuels and structural materials especially for the prototype reactor MONJU. With respect to the prototype reactor MONJU, the installation permit was issued on May 27, 1983, from the prime minister, and the contracts of the first stage between PNC and fabricators were made recently. At the same time, almost all the licenses of preparatory construction works were issued by March 1983, and preparatory construction works were started in April 1983. On the other hand, conceptual design of a demonstration reactor is now under way in a close cooperation with concerned authorities and utilities, as well as investigations of the way of conducting necessary research and development

  15. Oxide dispersion strengthened ferritic steels: a basic research joint program in France

    Energy Technology Data Exchange (ETDEWEB)

    Boutard, J.-L., E-mail: jean-louis.boutard@cea.fr [Cabinet du Haut-Commissaire, CEA/Saclay, 91191 Gif sur Yvette Cedex (France); Badjeck, V. [LPS, UMR CNRS 8502, Building 510, Université Paris-Sud 11, 91405 Orsay Cedex (France); Barguet, L. [LAUM, UMR CNRS 6613, Building IAM – UFR Sciences, Avenue O. Messiaen, 72085 Le Mans Cedex 9 (France); Barouh, C. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); Bhattacharya, A. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); CSNSM, UMR CNRS 8609, Université Paris-Sud 11, Buildings 104 and 108, 91405 Orsay Campus (France); Colignon, Y. [IM2NP, UMR CNRS 7334, Case 142, Faculté des Sciences, Campus de Saint Jérôme, Aix Marseille Université, 13397 Marseille Cedex 20 (France); Hatzoglou, C. [GPM, UMR CNRS 6634, Technopôle du Madrillet, Avenue de l’Université, BP12, 76801 Saint Etienne du Rouvray Cedex (France); Loyer-Prost, M. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France); Rouffié, A.L. [DMN/SRMA, CEA/Saclay, Building 455, 91191 Gif sur Yvette Cedex (France); Sallez, N. [SIMAP, UMR CNRS 5266, INPG, Domaine Universitaire, 1130 rue de la Piscine, BP75, 38402 Saint Martin d’Hères Cedex (France); Salmon-Legagneur, H. [DMN/SRMA, CEA/Saclay, Building 455, 91191 Gif sur Yvette Cedex (France); Schuler, T. [DMN/SRMP, CEA/Saclay, Building 520, 91191 Gif sur Yvette Cedex (France)

    2014-12-15

    AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9–14% Cr ferritic–martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical–chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α′ unmixing.

  16. Oxide dispersion strengthened ferritic steels: a basic research joint program in France

    Science.gov (United States)

    Boutard, J.-L.; Badjeck, V.; Barguet, L.; Barouh, C.; Bhattacharya, A.; Colignon, Y.; Hatzoglou, C.; Loyer-Prost, M.; Rouffié, A. L.; Sallez, N.; Salmon-Legagneur, H.; Schuler, T.

    2014-12-01

    AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9-14% Cr ferritic-martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical-chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α‧ unmixing.

  17. Oxide dispersion strengthened ferritic steels: a basic research joint program in France

    International Nuclear Information System (INIS)

    Boutard, J.-L.; Badjeck, V.; Barguet, L.; Barouh, C.; Bhattacharya, A.; Colignon, Y.; Hatzoglou, C.; Loyer-Prost, M.; Rouffié, A.L.; Sallez, N.; Salmon-Legagneur, H.; Schuler, T.

    2014-01-01

    AREVA, CEA, CNRS, EDF and Mécachrome are funding a joint program of basic research on Oxide Dispersion Strengthened Steels (ODISSEE), in support to the development of oxide dispersion strengthened 9–14% Cr ferritic–martensitic steels for the fuel element cladding of future Sodium-cooled fast neutron reactors. The selected objectives and the results obtained so far will be presented concerning (i) physical–chemical characterisation of the nano-clusters as a function of ball-milling process, metallurgical conditions and irradiation, (ii) meso-scale understanding of failure mechanisms under dynamic loading and creep, and, (iii) kinetic modelling of nano-clusters nucleation and α/α′ unmixing

  18. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    Khedr, H.

    2008-01-01

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

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

    International Nuclear Information System (INIS)

    Loika, E.F.

    1994-01-01

    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

  20. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  1. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA (United States); Biancheria, A

    1977-04-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  2. Experimental study of nitrogen oxides in the IRT-M reactor

    International Nuclear Information System (INIS)

    Brazovskij, I.I.; Doroshevich, V.N.; Gvozdev, A.A.; Nesterenko, V.B.; Trubnikov, V.P.

    1982-01-01

    A critical review of different approaches to the radiolysis study of nitrogen oxide under mixed radiation conditions of a nuclear reactor was presented. Loop reactor piant opereted following gas-liquid cycle. It was shown in the process of long experiment in the operating conditions that irreversible radiation-thermal decomposition of the coolant increases little with temperature and pressure and radioactivity of the coolant and thermophysical equipment was moderate. Numerous kinetic experiments were conducted on the ampoule plant wherein all coolant existed in the zone of ionizing radiation effect. Initial pressure in the ampoule plant was set in the range of 0.1-16 MPa, depending on conditions of the experiment, and temperature 200-500 deg C. Dosimetry of the ampoule was carried out by the radiolysis of nitrogen monoxide. The analysis of the radiolysis products was conducted utilizing gas chromatography method, coolant vapours were removed in the process of low-temperature condensation under - 70 deg C

  3. Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel

    International Nuclear Information System (INIS)

    Boltax, A.; Biancheria, A.

    1977-01-01

    Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction (FCMI) effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. (author)

  4. Oxidation damage evaluation by non-destructive method for graphite components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Tada, Tatsuya; Sumita, Junya; Sawa, Kazuhiro

    2008-01-01

    To develop non-destructive evaluation methods for oxidation damage on graphite components in High Temperature Gas-cooled Reactors (HTGRs), the applicability of ultrasonic wave and micro-indentation methods were investigated. Candidate graphites, IG-110 and IG-430, for core components of Very High Temperature Reactor (VHTR) were used in this study. These graphites were oxidized uniformly by air at 500degC. The following results were obtained from this study. (1) Ultrasonic wave velocities with 1 MHz can be expressed empirically by exponential formulas to burn-off, oxidation weight loss. (2) The porous condition of the oxidized graphite could be evaluated with wave propagation analysis with a wave-pore interaction model. It is important to consider the non-uniformity of oxidized porous condition. (3) Micro-indentation method is expected to determine the local oxidation damage. It is necessary to assess the variation of the test data. (author)

  5. High temperature CO2 capture using calcium oxide sorbent in a fixed-bed reactor

    International Nuclear Information System (INIS)

    Dou Binlin; Song Yongchen; Liu Yingguang; Feng Cong

    2010-01-01

    The gas-solid reaction and breakthrough curve of CO 2 capture using calcium oxide sorbent at high temperature in a fixed-bed reactor are of great importance, and being influenced by a number of factors makes the characterization and prediction of these a difficult problem. In this study, the operating parameters on reaction between solid sorbent and CO 2 gas at high temperature were investigated. The results of the breakthrough curves showed that calcium oxide sorbent in the fixed-bed reactor was capable of reducing the CO 2 level to near zero level with the steam of 10 vol%, and the sorbent in CaO mixed with MgO of 40 wt% had extremely low capacity for CO 2 capture at 550 deg. C. Calcium oxide sorbent after reaction can be easily regenerated at 900 deg. C by pure N 2 flow. The experimental data were analyzed by shrinking core model, and the results showed reaction rates of both fresh and regeneration sorbents with CO 2 were controlled by a combination of the surface chemical reaction and diffusion of product layer.

  6. Integrated bio-oxidation and adsorptive filtration reactor for removal of arsenic from wastewater.

    Science.gov (United States)

    Kamde, Kalyani; Dahake, Rashmi; Pandey, R A; Bansiwal, Amit

    2018-01-08

    Recently, removal of arsenic from different industrial effluent discharged using simple, efficient and low-cost technique has been widely considered. In this study, removal of arsenic (As) from real wastewater has been studied employing modified bio-oxidation followed by adsorptive filtration method in a novel continuous flow through the reactor. This method includes biological oxidation of ferrous to ferric ions by immobilized Acidothiobacillus ferrooxidans bacteria on granulated activated carbon (GAC) in fixed bed bio-column reactor with the adsorptive filtration unit. Removal efficiency was optimized regarding the initial flow rate of media and ferrous ions concentration. Synthetic wastewater sample having different heavy metal ions such as Arsenic (As), Cobalt (Co), Chromium (Cr), Copper (Cu), Iron (Fe), Lead (Pb) and Manganese (Mn) were also used in the study. The structural and surface changes occurring after the treatment process were scrutinized using FT-IR and Scanning Electron Microscopy (SEM) analysis. The finding showed that not only arsenic can be removed considerably in the bioreactor system, but also removing efficiency was much more (oxidation with adsorptive filtration method improves the removal efficiency of arsenic and other heavy metal ions in wastewater sample.

  7. Pilot program: NRC severe reactor accident incident response training manual: Severe reactor accident overview

    International Nuclear Information System (INIS)

    McKenna, T.J.; Martin, J.A.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

    1987-02-01

    This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Severe Reactor Accident Overview is the second in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident assesment. Each volume serves, respectively, as the text for a course of instruction in a series of courses. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

  8. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

    The main R and D results of Japanese activities are summarized as follows: (1) the experimental 140 MW(th) sodium cooled fast reactor 'Joyo' provided abundant experimental data and excellent operational records, attaining more than 50,000 hours of operation since its first criticality in 1977; (2) the prototype 280 MW(e) fast reactor 'Monju' reached initial criticality on 5 April 1994; presently Monju is under the cold shutdown state because of secondary sodium leak on 8 December 1995, and multiple cause investigations of the sodium leak are being performed; (3) the Japan Atomic Power Company is promoting design studies for demonstration fast reactor (DFBR) with a power output of 600 MW(e) and R and D for DFBR are being conducted under the cooperation of governmental and private sectors. (author)

  9. A review of fast reactor program in Japan - April 1983

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1983-01-01

    The fast breeder reactor development project in Japan has been in progress during the past twelve months and will be continued in the next fiscal year, from April 1983 through March 1984, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1982. Concerning the experimental fast reactor, JOYO, all the scheduled testings and operations were completed by the end of 1981 and from the beginning of 1982 the change-out work from Mark-I core to Mark-II core has been continued for 11 months. The initial criticality on the Mark-II core was achieved on 22 Nov. 1982 and after 3 months low power physics tests the reactor power was raised up to 100% (100 MWt) in the middle of March 1983. With respect to the prototype reactor MONJU, progress toward construction has been made and the licensing of the second step will be completed in the first half of 1983. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  10. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    International Nuclear Information System (INIS)

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions

  11. French studies and research program in pressurized water reactor safety

    International Nuclear Information System (INIS)

    Duco, J.

    1986-06-01

    The aim of researches developed now in France on water reactor safety is to obtain means and knowledge allowing to control accidental situations, including severe situations beyond design basis accidents. The main studies and researches concerning water reactors and described in this report are the following ones: core cooling accident and prevention of severe accidents, fuel behavior in accidental situation, behavior of the containment building, fission product transfer and releases in case of accident, problems related to equipment aging, and, methodology of risk analysis and ''human factor'' studies. Most of these studies follow an analytic approach of phenomena [fr

  12. The industry/EPRI advanced light water reactor program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Sugnet, W.R.; Bilan, W.J.

    1986-01-01

    For the United States nuclear power industry to remain viable, it must be prepared to meet the expected need for new generating capacity in the late 1990s with an improved reactor system. The best hope of meeting this requirement is with evolutionary changes in current LWR systems through system simplification and reevaluation of safety and operational design margins. The grid characteristics and the difficulty in raising capital for large projects indicate that smaller light water reactors (400 to 600 MWe) may play an important role the next generation

  13. Overview of U.S. Fast Reactor Technology Program

    International Nuclear Information System (INIS)

    Hill, Robert

    2013-01-01

    • Concept development studies guide R&D tasks by evaluating system impact for broad variety of technology options: – Small-scale facilities for R&D on key technology; – No near-term plan for demonstration reactor. • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction): – Advanced Structural Materials; – Advanced Energy Conversion; – Advanced Modeling and Simulation. • Other R&D is conducted to address known technology challenges: – Safety and Licensing; – Fuels Development; – Undersodium Viewing

  14. Investigations of the natural fission reactor program. Progress report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Cowan, G.A.; Norris, A.E.

    1978-10-01

    The U.S. study of the Oklo natural reactor began in 1973 with the principal objectives of understanding the processes that produced the reactor and that led to the retention of many of its products. Major facets of the program have been the chemical separation and mass spectrometric analysis of the reactor components and products, the petrological and mineralogical examination of samples taken from the reactor zones, and an interdisciplinary modeling of possible processes consistent with reactor physics, geophysics, and geochemistry. Most of the past work has been on samples taken within the reactor zones. Presently, these studies give greater emphasis to the measurement of mobile products in additional suites of samples collected peripherally and ''downstream'' from the reactor zones. This report summarizes the current status of research and the views of U.S. investigators, with particular reference to the extensive work of the French scientists, concerning the main features of the Oklo natural fission reactor. Also mentioned briefly is the U.S. search for natural fission reactors at other locations

  15. Enhanced xylene removal by photocatalytic oxidation using fiber-illuminated honeycomb reactor at ppb level

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Yi-Ting [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Yu, Yi-Hui [Department of Civil Engineering, National Taiwan University, Taipei 106, Taiwan (China); Nguyen, Van-Huy [Department of Chemical Engineering, National Taiwan University of Science and Technology, Taipei 10617, Taiwan (China); Lu, Kung-Te [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Wu, Jeffrey Chi-Sheng, E-mail: cswu@ntu.edu.tw [Department of Chemical Engineering, National Taiwan University, Taipei 10617, Taiwan (China); Chang, Luh-Maan [Department of Civil Engineering, National Taiwan University, Taipei 106, Taiwan (China); Kuo, Chi-Wen [Taiwan Semiconductor Manufacturing Company, Hsinchu 30078, Taiwan (China)

    2013-11-15

    Graphical abstract: We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. -- Highlights: • The combination of optical fiber and honeycomb significantly enhanced the performance of VOCs photodegradation. • The removal efficiency of m-xylene is enhanced to 96.5% as compared to 22.0% for UV irradiation alone. • Fiber-illuminated honeycomb reactor is the first step toward an industrial-scale technology on the removal of xylene. -- Abstract: The removal of volatile organic compounds (VOCs) at ppb level is one of the most critical challenges in clean rooms for the semiconductor industry. Photocatalytic oxidation is an innovative and promising technology for ppb-level VOCs degradation. We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. Using the FIHR with Mn-TiO{sub 2} photocatalyst not only increased the m-xylene removal efficiency, but also increased the CO{sub 2} selectivity. Interestingly, Mn-TiO{sub 2} in FIHR also showed a very good reusability, 93% removal efficiency was still achieved in 72-h in reaction. Thus, the FIHR gave very high removal efficiency for xylene at ppb level under room temperature. The FIHR has great potential application in the clean room for the air purification system in the future.

  16. Enhanced xylene removal by photocatalytic oxidation using fiber-illuminated honeycomb reactor at ppb level

    International Nuclear Information System (INIS)

    Wu, Yi-Ting; Yu, Yi-Hui; Nguyen, Van-Huy; Lu, Kung-Te; Wu, Jeffrey Chi-Sheng; Chang, Luh-Maan; Kuo, Chi-Wen

    2013-01-01

    Graphical abstract: We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. -- Highlights: • The combination of optical fiber and honeycomb significantly enhanced the performance of VOCs photodegradation. • The removal efficiency of m-xylene is enhanced to 96.5% as compared to 22.0% for UV irradiation alone. • Fiber-illuminated honeycomb reactor is the first step toward an industrial-scale technology on the removal of xylene. -- Abstract: The removal of volatile organic compounds (VOCs) at ppb level is one of the most critical challenges in clean rooms for the semiconductor industry. Photocatalytic oxidation is an innovative and promising technology for ppb-level VOCs degradation. We have designed a fiber-illuminated honeycomb reactor (FIHR) in which the removal efficiency of m-xylene is significantly enhanced to 96.5% as compared to 22.0% for UV irradiation only. The results indicate that photocatalysts not only play the role to substantially oxidize m-xylene, but also alter the chemical properties of xylene under UV illumination. Using the FIHR with Mn-TiO 2 photocatalyst not only increased the m-xylene removal efficiency, but also increased the CO 2 selectivity. Interestingly, Mn-TiO 2 in FIHR also showed a very good reusability, 93% removal efficiency was still achieved in 72-h in reaction. Thus, the FIHR gave very high removal efficiency for xylene at ppb level under room temperature. The FIHR has great potential application in the clean room for the air purification system in the future

  17. The Enhancement of the Selectivity of Complex Reactions by a Catalytic Membrane Reactor -Ethylene Oxidation Over a Ag Catalyst Supported in a Ceramic Membrane-

    OpenAIRE

    馮, 臨; 小林, 正義; Lin, FENG; Masayoshi, KOBAYASHI

    1991-01-01

    This research demonstrated that, using a membrane reactor consisting of a tubular, microporous, glass-ceramic membrane, it is possible to achieve selective oxidation of ethylene to ethylene oxide with an Ag catalyst. In experiments which a reaction temperature range of 115 to 300℃ and a contact time of 1.5 to 5 seconds, resulting data illustrated the following characteristics of this membrane reactor : 1) compared with a classic tubular reactor, the selectivity of ethylene oxide is increased ...

  18. Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

  19. Reprocessing fuel from the Southwest Experimental Fast Oxide Reactor at the Savannah River Plant

    International Nuclear Information System (INIS)

    Gray, L.W.; Campbell, T.G.

    1985-11-01

    The irradiated fuel, reject fuel tubes, and fuel fabrication scrap from the Southwest Experimental Fast Oxide Reactor (SEFOR) were transferred to the Savannah River Plant (SRP) for uranium and plutonium recovery. The unirradiated material was declad and dissolved at SRP; dissolution was accomplished in concentrated nitric acid without the addition of fluoride. The irradiated fuel was declad at Atomics International and repacked in aluminum. The fuel and aluminum cans were dissolved at SRP using nitric acid catalyzed by mercuric nitrate. As this fuel was dissolved in nongeometrically favorable tanks, boron was used as a soluble neutron poison

  20. Experimental PIV and CFD studies of UV-peroxide advanced oxidation reactors for water treatment

    International Nuclear Information System (INIS)

    Sozzi, A.; Taghipour, F.

    2004-01-01

    An experimental and numerical study of the flow characteristics in an annular UV reactor, as used for drinking water disinfection or Advanced Oxidation Processes, was carried out using Particle Image Velocimetry (PIV) and Computational Fluid Dynamics (CFD). The influence of different turbulence models and mesh structures on the CFD results was investigated. By qualitative and quantitative comparison of CFD and PIV experimental data, it was shown that the Realizable k-e- turbulence model is best suited for simulating the hydrodynamics of this geometry. (author)

  1. In situ secondary organic aerosol formation from ambient pine forest air using an oxidation flow reactor

    Science.gov (United States)

    Palm, Brett B.; Campuzano-Jost, Pedro; Ortega, Amber M.; Day, Douglas A.; Kaser, Lisa; Jud, Werner; Karl, Thomas; Hansel, Armin; Hunter, James F.; Cross, Eben S.; Kroll, Jesse H.; Peng, Zhe; Brune, William H.; Jimenez, Jose L.

    2016-03-01

    An oxidation flow reactor (OFR) is a vessel inside which the concentration of a chosen oxidant can be increased for the purpose of studying SOA formation and aging by that oxidant. During the BEACHON-RoMBAS (Bio-hydro-atmosphere interactions of Energy, Aerosols, Carbon, H2O, Organics & Nitrogen-Rocky Mountain Biogenic Aerosol Study) field campaign, ambient pine forest air was oxidized by OH radicals in an OFR to measure the amount of SOA that could be formed from the real mix of ambient SOA precursor gases, and how that amount changed with time as precursors changed. High OH concentrations and short residence times allowed for semicontinuous cycling through a large range of OH exposures ranging from hours to weeks of equivalent (eq.) atmospheric aging. A simple model is derived and used to account for the relative timescales of condensation of low-volatility organic compounds (LVOCs) onto particles; condensational loss to the walls; and further reaction to produce volatile, non-condensing fragmentation products. More SOA production was observed in the OFR at nighttime (average 3 µg m-3 when LVOC fate corrected) compared to daytime (average 0.9 µg m-3 when LVOC fate corrected), with maximum formation observed at 0.4-1.5 eq. days of photochemical aging. SOA formation followed a similar diurnal pattern to monoterpenes, sesquiterpenes, and toluene+p-cymene concentrations, including a substantial increase just after sunrise at 07:00 local time. Higher photochemical aging (> 10 eq. days) led to a decrease in new SOA formation and a loss of preexisting OA due to heterogeneous oxidation followed by fragmentation and volatilization. When comparing two different commonly used methods of OH production in OFRs (OFR185 and OFR254-70), similar amounts of SOA formation were observed. We recommend the OFR185 mode for future forest studies. Concurrent gas-phase measurements of air after OH oxidation illustrate the decay of primary VOCs, production of small oxidized organic

  2. In situ secondary organic aerosol formation from ambient pine forest air using an oxidation flow reactor

    Directory of Open Access Journals (Sweden)

    B. B. Palm

    2016-03-01

    Full Text Available An oxidation flow reactor (OFR is a vessel inside which the concentration of a chosen oxidant can be increased for the purpose of studying SOA formation and aging by that oxidant. During the BEACHON-RoMBAS (Bio-hydro-atmosphere interactions of Energy, Aerosols, Carbon, H2O, Organics & Nitrogen–Rocky Mountain Biogenic Aerosol Study field campaign, ambient pine forest air was oxidized by OH radicals in an OFR to measure the amount of SOA that could be formed from the real mix of ambient SOA precursor gases, and how that amount changed with time as precursors changed. High OH concentrations and short residence times allowed for semicontinuous cycling through a large range of OH exposures ranging from hours to weeks of equivalent (eq. atmospheric aging. A simple model is derived and used to account for the relative timescales of condensation of low-volatility organic compounds (LVOCs onto particles; condensational loss to the walls; and further reaction to produce volatile, non-condensing fragmentation products. More SOA production was observed in the OFR at nighttime (average 3 µg m−3 when LVOC fate corrected compared to daytime (average 0.9 µg m−3 when LVOC fate corrected, with maximum formation observed at 0.4–1.5 eq. days of photochemical aging. SOA formation followed a similar diurnal pattern to monoterpenes, sesquiterpenes, and toluene+p-cymene concentrations, including a substantial increase just after sunrise at 07:00 local time. Higher photochemical aging (> 10 eq. days led to a decrease in new SOA formation and a loss of preexisting OA due to heterogeneous oxidation followed by fragmentation and volatilization. When comparing two different commonly used methods of OH production in OFRs (OFR185 and OFR254-70, similar amounts of SOA formation were observed. We recommend the OFR185 mode for future forest studies. Concurrent gas-phase measurements of air after OH oxidation illustrate the decay of primary VOCs, production

  3. Cross-disciplinary research programs at the Cornell TRIGA reactor

    International Nuclear Information System (INIS)

    Clark, D.D.

    1995-01-01

    This paper describes cross-disciplinary research efforts at the Cornell TRIGA reactor. A new graduate laboratory course for nonspecialists was developed which brought in graduate students from many fields, and a weekly or bimonthly nuclear methods seminars are being held to describe research methods, sample preparation, irradiation, etc

  4. Application of microprocessor based controller in the Breeder Reactor Program

    International Nuclear Information System (INIS)

    Messick, N.C.; Lukas, M.P.

    1985-01-01

    This paper treats Argonne National Laboratory's experience using microprocessor based controllers presently in use on several control loops within the EBR-II reactor facility as well as tests being performed by these controllers. Also included is a discussion of the expandability, modularity, range of capabilities and higher level functions possible using such equipment

  5. NEPTUNIX, a general program of simulation applied to nuclear reactors

    International Nuclear Information System (INIS)

    Bonnemay, A.; Dansac Bon, V.

    1978-01-01

    Most simulation languages admit an incremental description and involve explicit integration algorithms. NEPTUNIX is a simulation language directly admitting algebraic differential equations under an implicit form, and it involves a very efficient implicit integration method with variable step and order. NEPTUNIX is a tool used for building large systems models in the field of nuclear reactors [fr

  6. Programmed elimination of neutronic poisons in nuclear reactors

    International Nuclear Information System (INIS)

    Perriere, G. de la

    1967-11-01

    This work deals with the use of salts of elements having a large neutron capture cross-section, so-called 'soluble poisons' which are dissolved in the moderating water to control the reactivity of heavy-water reactors, and more particularly to compensate the xenon effect in the reactor EL 4. The report describes the controlled elimination of these poisons by fixation on ion-exchange resins. The poisons considered are lithium-6, cadmium and gadolinium in the sulphate form, and boron as boric acid. The thermodynamic and kinetic constants of the ion-exchange reactions were first determined and a study was then made of the fixation of these compounds in beds of small-calibre resins placed in columns. Lithium-6 is the poison which is most easily applicable to compensate the xenon effect in the reactor EL 4. It can be eliminated rapidly and completely from heavy water, and its use does not lead to supplementary problems of protection against the gamma radiation of the reactor circuits. (author) [fr

  7. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

  8. Reactive turbulent flow CFD study in supercritical water oxidation process: application to a stirred double shell reactor

    International Nuclear Information System (INIS)

    Moussiere, S.

    2006-12-01

    Supercritical water oxidation is an innovative process to treat organic liquid waste which uses supercritical water properties to mix efficiency the oxidant and the organic compounds. The reactor is a stirred double shell reactor. In the step of adaptation to nuclear constraints, the computational fluid dynamic modeling is a good tool to know required temperature field in the reactor for safety analysis. Firstly, the CFD modeling of tubular reactor confirms the hypothesis of an incompressible fluid and the use of k-w turbulence model to represent the hydrodynamic. Moreover, the EDC model is as efficiency as the kinetic to compute the reaction rate in this reactor. Secondly, the study of turbulent flow in the double shell reactor confirms the use of 2D axisymmetric geometry instead of 3D geometry to compute heat transfer. Moreover, this study reports that water-air mixing is not in single phase. The reactive turbulent flow is well represented by EDC model after adaptation of initial conditions. The reaction rate in supercritical water oxidation reactor is mainly controlled by the mixing. (author)

  9. Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China); Lee, Hyung-Sool [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West Waterloo, Ontario, Canada N2L 3G1 (Canada); Wang, Ai-Jie, E-mail: waj0578@hit.edu.cn [State Key Laboratory of Urban Water Resource and Environment, Harbin Institute of Technology, No. 202 Haihe Road, Harbin 150090 (China)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. Black-Right-Pointing-Pointer Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). Black-Right-Pointing-Pointer PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. Black-Right-Pointing-Pointer The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 {+-} 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m{sup -3} d{sup -1}) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m{sup -3} d{sup -1} (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

  10. Azo dye removal in a membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor

    International Nuclear Information System (INIS)

    Cui, Dan; Guo, Yu-Qi; Cheng, Hao-Yi; Liang, Bin; Kong, Fan-Ying; Lee, Hyung-Sool; Wang, Ai-Jie

    2012-01-01

    Highlights: ► A membrane-free up-flow biocatalyzed electrolysis reactor coupled with an aerobic bio-contact oxidation reactor was developed. ► Alizarin Yellow R as the mode of azo dyes was efficiently converted to p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA). ► PPD and 5-ASA were further oxidized in a bio-contact oxidation reactor. ► The mechanism of UBER for azo dye removal was discussed. - Abstract: Azo dyes that consist of a large quantity of dye wastewater are toxic and persistent to biodegradation, while they should be removed before being discharged to water body. In this study, Alizarin Yellow R (AYR) as a model azo dye was decolorized in a combined bio-system of membrane-free, continuous up-flow bio-catalyzed electrolysis reactor (UBER) and subsequent aerobic bio-contact oxidation reactor (ABOR). With the supply of external power source 0.5 V in the UBER, AYR decolorization efficiency increased up to 94.8 ± 1.5%. Products formation efficiencies of p-phenylenediamine (PPD) and 5-aminosalicylic acid (5-ASA) were above 90% and 60%, respectively. Electron recovery efficiency based on AYR removal in cathode zone was nearly 100% at HRTs longer than 6 h. Relatively high concentration of AYR accumulated at higher AYR loading rates (>780 g m −3 d −1 ) likely inhibited acetate oxidation of anode-respiring bacteria on the anode, which decreased current density in the UBER; optimal AYR loading rate for the UBER was 680 g m −3 d −1 (HRT 2.5 h). The subsequent ABOR further improved effluent quality. Overall the Chroma decreased from 320 times to 80 times in the combined bio-system to meet the textile wastewater discharge standard II in China.

  11. The in-core experimental program at the MIT Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kohse, G.E.; Hu, L-W., E-mail: kohse@mit.edu [Massachusetts Inst. of Technology, Nuclear Reactor Lab., Cambridge, Massachusetts (United States)

    2014-07-01

    This paper describes the program of in-core experiments at the Massachusetts Institute of Technology Research Reactor (MITR), a 6 MW research reactor. The MITR has a neutron flux and spectrum similar to those in water-cooled power reactors and therefore provides a useful test environment for materials and fuels research. In-core facilities include: a water loop operating at pressurized water or boiling water reactor conditions, an inert gas irradiation facility operating at temperature up to 850 {sup o}C and special purpose facilities including fuel irradiation experiments. Recent and ongoing tests include: water loop investigations of corrosion and thermal and mechanical property evolution of SiC/SiC composites for fuel cladding, irradiation of advanced materials and in-core sensors at elevated temperatures, irradiation in molten fluoride salt at 700 {sup o}C of metal alloy, graphite and composite materials for power reactor applications and instrumented irradiations of metal-bonded hydride fuel. (author)

  12. A dedicated program for the extended longevity of research and training reactors

    International Nuclear Information System (INIS)

    Carriveau, G.W.

    1992-01-01

    In the past 49 years, over 555 research and training reactors have been in operation, with approximately 325 currently in service. The age distribution of operating research reactors shows that the average age is about 24 years; about 74% are 20 years old or older and about 33% are 30 years old or older. This group of reactors represents a very large investment in capital expense with replacement costs in 1990's prices much higher than when they were originally constructed. Furthermore, decommissioning costs may be much greater than the original investments. General Atomics has been directly involved for the better part of the nearly fifty year history of research and training reactors. This paper will describe a General Atomics program illustrating a dedicated commitment to the full service support of extended and improved use for all types of research and training reactors. (author)

  13. A continuing success - The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Mustin, Tracy P.; Clapper, Maureen; Reilly, Jill E.

    2000-01-01

    The United States Department of Energy, in consultation with the Department of State, adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program, established under this policy, has completed 16 spent fuel shipments. 2,651 material test reactor (MTR) assemblies, one Slowpoke core containing less than 1 kilogram of U.S.-origin enriched uranium, 824 Training, Research, Isotope, General Atomic (TRIGA) rods, and 267 TRIGA pins from research reactors around the world have been shipped to the United States so far under this program. As the FRR SNF Acceptance Program progresses into the fifth year of implementation, a second U.S. cross country shipment has been completed, as well as a second overland truck shipment from Canada. Both the cross country shipment and the Canadian shipment were safely and successfully completed, increasing our knowledge and experience in these types of shipments. In addition, two other shipments were completed since last year's RERTR meeting. Other program activities since the last meeting included: taking pre-emptive steps to avoid license amendment pitfalls/showstoppers for spent fuel casks, publication of a revision to the Record of Decision allowing up to 16 casks per ocean going vessel, and the issuance of a cable to 16 of the 41 eligible countries reminding their governments and the reactor operators that the U.S.-origin uranium in their research reactors may be eligible for return to the United States under the Acceptance Program and urging them to begin discussions on shipping schedules. The FRR SNF program has also supported the Department's implementation of the competitive pricing policy for uranium and resumption of shipments of fresh uranium for fabrication into assemblies for research reactors. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues

  14. Fiscal year 1998 multi-year work plan. Advanced reactors transition program

    International Nuclear Information System (INIS)

    Gantt, D.A.

    1997-01-01

    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. Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

    International Nuclear Information System (INIS)

    Rice, R.E.

    1976-09-01

    Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments

  16. NRC review of passive reactor design certification testing programs: Overview, progress, and regulatory perspective

    Energy Technology Data Exchange (ETDEWEB)

    Levin, A.E.

    1995-09-01

    New reactor designs, employing passive safety systems, are currently under development by reactor vendors for certification under the U.S. Nuclear Regulatory Commission`s (NRC`s) design certification rule. The vendors have established testing programs to support the certification of the passive designs, to meet regulatory requirements for demonstration of passive safety system performance. The NRC has, therefore, developed a process for the review of the vendors` testing programs and for incorporation of the results of those reviews into the safety evaluations for the passive plants. This paper discusses progress in the test program reviews, and also addresses unique regulatory aspects of those reviews.

  17. ARIES Oxide Production Program Annual Report - FY14

    Energy Technology Data Exchange (ETDEWEB)

    Kelley, Evelyn A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dinehart, Steven Mark [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-01

    A summary of the major accomplishments (September), milestones, financial summary, project performance and issues facing the ARIES Oxide Production Program at the close of FY14 is presented in this Executive Summary. Annual accomplishments are summarized in the body of the report.

  18. Optimization programs for reactor core fuel loading exhibiting reduced neutron leakage

    International Nuclear Information System (INIS)

    Darilek, P.

    1991-01-01

    The program MAXIM was developed for the optimization of the fuel loading of WWER-440 reactors. It enables the reactor core reactivity to be maximized by modifying the arrangement of the fuel assemblies. The procedure is divided into three steps. The first step includes the passage from the three-dimensional model of the reactor core to the two-dimensional model. In the second step, the solution to the problem is sought assuming that the multiplying properties, or the reactivity in the zones of the core, vary continuously. In the third step, parameters of actual fuel assemblies are inserted in the ''continuous'' solution obtained. Combined with the program PROPAL for a detailed refinement of the loading, the program MAXIM forms a basis for the development of programs for the optimization of fuel loading with burnable poisons. (Z.M.). 16 refs

  19. Aeration Strategies To Mitigate Nitrous Oxide Emissions from Single-Stage Nitritation/Anammox Reactors

    DEFF Research Database (Denmark)

    Domingo Felez, Carlos; Mutlu, A. Gizem; Jensen, Marlene Mark

    2014-01-01

    Autotrophic nitrogen removal is regarded as a resource efficient process to manage nitrogen-rich residual streams. However, nitrous oxide emissions of these processes are poorly documented and strategies to mitigate emissions unknown. In this study, two sequencing batch reactors performing single...... was noted when the duration of aeration was increased while decreasing air flow rate (10.9 +/- 3.2% Delta N2O/Delta TN). The extant ammonium oxidation activity (mgNH(4)(+)-N/gVSS.min) positively correlated with the specific N2O production rate (mgN(2)O-N/gVSS.min) of the systems. Operating under conditions......-stage nitritation/anammox were operated under different aeration strategies, gradually adjusted over six months. At constant but limiting oxygen loading, synthetic reject water was fed (0.75g-N/L.d) and high nitrogen removal efficiencies (83 +/- 5 and 88 +/- 2%) obtained. Dynamics of liquid phase nitrous (N2O...

  20. Development of inspection and maintenance program for reactor and reactivity control units in HANARO

    International Nuclear Information System (INIS)

    Cho, Yeong-Garp

    1998-01-01

    This paper summarizes the overall program for inspection and maintenance of reactor structure and Reactivity Control Units (RCU) of HANARO during lifetime. The long-term plan for in-service inspection is introduced in the viewpoint of the structural integrity of reactor and RCU, and the operability of RCU mechanism. This program includes the list of components to be inspected, the schedule of inspection and maintenance, and the development of special tools and test rig that are required for the remote inspection and maintenance of reactor and RCU components. Preliminary results of the evaluation on the lifetime of RCU components are summarized based on the operation history since the installation of reactor. A test rig will be designed and constructed for the purposes of verifying the prolonged lifetime of RCU components being used, the performance of special tools, and the rehearsal of maintenance work as well. (author)

  1. Progress in space nuclear reactor power systems technology development - The SP-100 program

    Science.gov (United States)

    Davis, H. S.

    1984-01-01

    Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

  2. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods, Stage 18

    International Nuclear Information System (INIS)

    Pazsit, Imre; Nam, Tran Hoai; Dykin, Victor; Jonsson, Anders

    2013-01-01

    This report constitutes Stage 18 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The objective of the research program is to contribute to the strategic research goal of competence and research capacity by building up competence within the Department of Nuclear Engineering at Chalmers University of Technology, regarding reactor physics, reactor dynamics and noise diagnostics. The purpose is also to contribute to the research goal of giving a basis for SSM's supervision by developing methods for identification and localization of perturbations in reactor cores. Results up to Stage 17 were reported in SKI and SSM reports, as listed in the report's summary

  3. Minority and female training programs at the Ford Nuclear Reactor, University of Michigan

    International Nuclear Information System (INIS)

    Burn, R.R.

    1992-01-01

    Nuclear power industry operations staffs are composed predominantly of white males because most of the personnel come from the nuclear submarine and surface branches of the U.S. Navy. The purpose of the minority and female training programs sponsored by the Ford Nuclear Reactor at the University of Michigan is to provide a path for minorities and women to enter the nuclear industry as operators, technicians, and, in the long term, as graduate engineers. The training programs are aimed at high school students, preferably juniors. While the training is directed toward operation of a nuclear reactor, it is equally applicable to careers in most other technical fields. It is hoped that some of the participants will remain at the Ford Nuclear Reactor as reactor operators, enter college, and obtain college degrees, after which they will enter the nuclear industry as graduate engineers

  4. Nuclear science. U.S. electricity needs and DOE's civilian reactor development program

    International Nuclear Information System (INIS)

    England-Joseph, Judy; Allen, Robert E. Jr.; Fitzgerald, Duane; Young, Edward E. Jr.; Leavens, William P.; Bell, Jacqueline

    1990-05-01

    Electricity projections developed by the North American Electric Reliability Council (NERC) appear to be the best available estimates of future U.S. electricity needs. NERC, which represents all segments of the utility industry, forecasts that before 1998 certain regions of the country, particularly in the more heavily populated eastern half of the United States, may experience shortfalls during summer peak demand periods. These forecasts considered the utility companies' plans, as of 1989, to meet electricity needs during the period; these plans include such measures as constructing additional generators and conducting demand management programs. Working closely with the nuclear industry, DOE is supporting the development of several reactor technologies to ensure that nuclear power remains a viable electricity supply option. In fiscal year 1990, DOE's Civilian Reactor Development Program was funded at $253 million. DOE is using these funds to support industry-led efforts to develop light water reactors (LWR), advanced liquid-metal reactors (LMR), and modular high-temperature gas-cooled reactors (MHTGR) that are safe, environmentally acceptable, and economically competitive. The utility company officials we spoke with, all of whom were in the Southeast, generally supported DOE's efforts in developing these technologies. However, most of the officials do not plan to purchase nuclear reactors until after 2000 because of the high costs of constructing nuclear reactors and current public opposition to nuclear power

  5. The development of the breeder reactor program of the German Federal Republic

    International Nuclear Information System (INIS)

    Haunschild, H.H.

    1984-01-01

    This article recapitulates the stages of the program which is in hand in the German Federal Republic concerning breeder reactors with fast neutrons. In particular, it describes in detail the causes for the delays in the construction of the demonstration reactor SRN 300 (Kalkar) for which the commissioning has now been fixed for 1987. It emphasizes the importance of the collaboration which exists in Europe for the establishment of the projects and the construction of fast neutron power stations, including the SNR 2 reactor; the construction of the latter could begin in the German Federal Republic at the end of the Eighties [fr

  6. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail

  7. Programming of computers for the protection system for Savannah River reactors

    International Nuclear Information System (INIS)

    Finley, R.H.

    1977-06-01

    The monitoring requirements for the SRP Safety Computers are shown. These fast response times coupled with the large number of analog inputs to be scanned imposed stringent program requirements. The system consists of two separate computers, each with its own inputs to monitor half the reactor positions. Either computer can provide the minimum required monitoring. The desired redundant monitoring is provided when both computers are on-line. If both computers are off-line, the reactor is automatically shut down

  8. Reactor R ampersand D programs tough to eliminate; just ask NRC staff

    International Nuclear Information System (INIS)

    Lane, E.

    1993-01-01

    Even if the Clinton administration succeeds in eliminating funding for the advanced liquid metal reactor (ALMR) and modular high-temperature gas reactor (MHTGR) in the fiscal year 1994 budget, it will not wipe out the programs entirely as shown by a recent exchange of letters between the Nuclear Regulatory Commission and the Energy Department. This article examines the political and bureaucratic maneuverings involved in the funding of nuclear power projects

  9. Using low-enriched uranium in research reactors: The RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1994-01-01

    The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of 99 Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program

  10. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    International Nuclear Information System (INIS)

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-01-01

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies

  11. Gas reactor international coope--ative program. Interim report: assessment of gas-cooled reactor economics

    Energy Technology Data Exchange (ETDEWEB)

    1979-08-01

    A computer analysis of domestic economic incentive is presented. Included are the sample computer data set for ten combinations of reprocessing and reactor assumptions; basic data set and computer output; higher uranium availability computer output; 50 percent higher GCR fabrication cost computer output; 50 percent higher GCR reprocessing cost computer output; year 1990 and year 2000 GCR introduction scenario computer outputs; 75 percent perceived capacity factor for PBR computer output; and capital cost of GCRs 1.2 times that of LWRs.

  12. SP-100 reactor disassembly remote handling test program

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Maiden, G.E.; Vader, D.P.

    1991-01-01

    This paper is presented as an overview of the remote handling equipment validation testing, which will be conducted before installation and use in the ground engineering test facility. This equipment will be used to defuel the SP-100 reactor core after removing it from the Test Assembly following nuclear testing. A series of full scale mock-up operational tests will be conducted at a Hanford Site facility to verify equipment design, operation, and capabilities

  13. Nitrous oxide production pathways in a partial nitritation-anammox reactor: Isotopic evidence for nitrous oxide production associated anaerobic ammonium oxidation?

    Science.gov (United States)

    Wunderlin, P.; Harris, E. J.; Joss, A.; Emmenegger, L.; Kipf, M.; Mohn, J.; Siegrist, H.

    2014-12-01

    Nitrous oxide (N2O) is a strong greenhouse gas and a major sink for stratospheric ozone. In biological wastewater treatment N2O can be produced via several pathways. This study investigates the dynamics of N2O emissions from a nitritation-anammox reactor, and links its interpretation to the nitrogen and oxygen isotopic signature of the emitted N2O. A 400-litre single-stage nitritation-anammox reactor was operated and continuously fed with digester liquid. The isotopic composition of N2O emissions was monitored online with quantum cascade laser absorption spectroscopy (QCLAS; Aerodyne Research, Inc.; Waechter et al., 2008). Dissolved ammonium and nitrate were monitored online (ISEmax, Endress + Hauser), while nitrite was measured with test strips (Nitrite-test 0-24mgN/l, Merck). Table 1. Summary of experiments conducted to understand N2O emissions Experimental conditions O2[mgO2/L] NO2-[mgN/L] NH4+[mgN/L] N2O/NH4+[%] Normal operation production pathway, which is hypothesized to be mediated by anammox activity (Figure 1). A less likely explanation is that the SP of N2O was increased by partial N2O reduction by heterotrophic denitrification. Various experiments were conducted to further investigate N2O formation pathways in the reactor. Our data reveal that N2O emissions increased when reactor operation was not ideal, for example when dissolved oxygen was too high (Table 1). SP measurements confirmed that these N2O peaks were due to enhanced nitrifier denitrification, generally related to nitrite build-up in the reactor (Figure 1; Table 1). Overall, process control via online N2O monitoring was confirmed to be an ideal method to detect imbalances in reactor operation and regulate aeration, to ensure optimal reactor conditions and minimise N2O emissions. ReferencesWaechter H. et al. (2008) Optics Express, 16: 9239-9244. Wunderlin, P et al. (2013) Environmental Science & Technology 47: 1339-1348.

  14. Diversity Profile of Microbes Associated with Anaerobic Sulfur Oxidation in an Upflow Anaerobic Sludge Blanket Reactor Treating Municipal Sewage

    Science.gov (United States)

    Aida, Azrina A.; Kuroda, Kyohei; Yamamoto, Masamitsu; Nakamura, Akinobu; Hatamoto, Masashi; Yamaguchi, Takashi

    2015-01-01

    We herein analyzed the diversity of microbes involved in anaerobic sulfur oxidation in an upflow anaerobic sludge blanket (UASB) reactor used for treating municipal sewage under low-temperature conditions. Anaerobic sulfur oxidation occurred in the absence of oxygen, with nitrite and nitrate as electron acceptors; however, reactor performance parameters demonstrated that anaerobic conditions were maintained. In order to gain insights into the underlying basis of anaerobic sulfur oxidation, the microbial diversity that exists in the UASB sludge was analyzed comprehensively to determine their identities and contribution to sulfur oxidation. Sludge samples were collected from the UASB reactor over a period of 2 years and used for bacterial 16S rRNA gene-based terminal restriction fragment length polymorphism (T-RFLP) and next-generation sequencing analyses. T-RFLP and sequencing results both showed that microbial community patterns changed markedly from day 537 onwards. Bacteria belonging to the genus Desulforhabdus within the phylum Proteobacteria and uncultured bacteria within the phylum Fusobacteria were the main groups observed during the period of anaerobic sulfur oxidation. Their abundance correlated with temperature, suggesting that these bacterial groups played roles in anaerobic sulfur oxidation in UASB reactors. PMID:25817585

  15. Calculation of gas-flow in plasma reactor for carbon partial oxidation

    Science.gov (United States)

    Bespala, Evgeny; Myshkin, Vyacheslav; Novoselov, Ivan; Pavliuk, Alexander; Makarevich, Semen; Bespala, Yuliya

    2018-03-01

    The paper discusses isotopic effects at carbon oxidation in low temperature non-equilibrium plasma at constant magnetic field. There is described routine of experiment and defined optimal parameters ensuring maximum enrichment factor at given electrophysical, gas-dynamic, and thermodymanical parameters. It has been demonstrated that at high-frequency generator capacity of 4 kW, supply frequency of 27 MHz and field density of 44 mT the concentration of paramagnetic heavy nuclei 13C in gaseous phase increases up to 1.78 % compared to 1.11 % for natural concentration. Authors explain isotopic effect decrease during plasmachemical separation induced by mixing gas flows enriched in different isotopes at the lack of product quench. With the help of modeling the motion of gas flows inside the plasma-chemical reactor based on numerical calculation of Navier-Stokes equation authors determine zones of gas mixing and cooling speed. To increase isotopic effects and proportion of 13C in gaseous phase it has been proposed to use quench in the form of Laval nozzle of refractory steel. The article represents results on calculation of optimal Laval Nozzle parameters for plasma-chemical reactor of chosen geometry of. There are also given dependences of quench time of products on pressure at the diffuser output and on critical section diameter. Authors determine the location of quench inside the plasma-chemical reactor in the paper.

  16. Refractory oxides for fusion reactor first walls, the effects of the reducing environment

    International Nuclear Information System (INIS)

    Hoffman, J.G.

    1979-01-01

    Of the several applications for refractory oxides in fusion reactor systems, the most demanding is that for the first wall. Some components in proximity of the first wall (possibly waveguides or flux breakers) will also be subjected to similar environments. Many parameters affect the ultimate usability of a particular material for reactor applications: electrical resistivity and dielectric breakdown if applicable, thermal conductivity, mechanical properties, and stability with respect to neutral molecular or atomic, or ionized fuel gases. All these properties can be affected by the radiation environment present in an operating power reactor. Temperatures up to 2000K may be expected for radiatively cooled first wall liners in some proposed designs although surface temperatures are appreciably lower (approximately 1000K) in other applications. The exact nature of the chemical environment is not defined even for the most well developed design concepts, but possible environments may be hypothesized; ambient neutral molecular and atomic species, bombardment by high energy charge exchange neutral atoms, direct ionic bombardment from stray ions, and plasma dumps from failure of the confinement system. Preliminary work has begun to more adequately define the extent of the problem and suggest approaches to engineering solutions

  17. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  18. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    International Nuclear Information System (INIS)

    Cowell, B.S.; Fisher, S.E.

    1999-01-01

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

  19. Methane oxidation in a landfill cover soil reactor: Changing of kinetic parameters and microorganism community structure.

    Science.gov (United States)

    Xing, Zhi L; Zhao, Tian T; Gao, Yan H; Yang, Xu; Liu, Shuai; Peng, Xu Y

    2017-02-23

    Changing of CH 4 oxidation potential and biological characteristics with CH 4 concentration was studied in a landfill cover soil reactor (LCSR). The maximum rate of CH 4 oxidation reached 32.40 mol d -1 m -2 by providing sufficient O 2 in the LCSR. The kinetic parameters of methane oxidation in landfill cover soil were obtained by fitting substrate diffusion and consumption model based on the concentration profile of CH 4 and O 2 . The values of [Formula: see text] (0.93-2.29%) and [Formula: see text] (140-524 nmol kg soil-DW -1 ·s -1 ) increased with CH 4 concentration (9.25-20.30%), while the values of [Formula: see text] (312.9-2.6%) and [Formula: see text] (1.3 × 10 -5 to 9.0 × 10 -3 nmol mL -1 h -1 ) were just the opposite. MiSeq pyrosequencing data revealed that Methylobacter (the relative abundance was decreased with height of LCSR) and Methylococcales_unclassified (the relative abundance was increased expect in H 80) became the key players after incubation with increasing CH 4 concentration. These findings provide information for assessing CH 4 oxidation potential and changing of biological characteristics in landfill cover soil.

  20. Quantities of Interest in Jet Stirred Reactor Oxidation of a High-Octane Gasoline

    KAUST Repository

    Chen, Bingjie

    2017-03-28

    This work examines the oxidation of a well-characterized, high-octane-number FACE (fuel for advanced combustion engines) F gasoline. Oxidation experiments were performed in a jet-stirred reactor (JSR) for FACE F gasoline under the following conditions: pressure, 10 bar; temperature, 530-1250 K; residence time, 0.7s; equivalence ratios, 0.5, 1.0, and 2.0. Detailed species profiles were achieved by identification and quantification from gas chromatography with mass spectrometry (GC-MS) and Fourier transform infrared spectrometry (FTIR). Four surrogates, with physical and chemical properties that mimic the real fuel properties, were used for simulations, with a detailed gasoline surrogate kinetic model. Fuel and species profiles were well-captured and-predicted by comparisons between experimental results and surrogate simulations. Further analysis was performed using a quantities of interest (QoI) approach to show the differences between experimental and simulation results and to evaluate the gasoline surrogate kinetic model. Analysis of the multicomponent surrogate kinetic model indicated that iso-octane and alkyl aromatic oxidation reactions had impact on species profiles in the high-temperature region;. however, the main production and consumption channels were related to smaller molecule reactions. The results presented here offer new insights into the oxidation chemistry of complex gasoline fuels and provide suggestions for the future development of surrogate kinetic models.

  1. Progress of the United States foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Huizenga, D.G.; Clapper, M.; Thrower, A.W.

    2002-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program has completed 23 shipments. Almost 5000 spent fuel assemblies from eligible research reactors throughout the world have been accepted into the United States under this program. Over the past year, another cross-country shipment of fuel was accomplished, as well as two additional shipments in the fourth quarter of calendar year 2001. These shipments attracted considerable safeguards oversight since they occurred post September 11. Recent guidance from the Nuclear Regulatory Commission (NRC) pertaining to security and safeguards issues deals directly with the transport of nuclear material. Since the Acceptance Program has consistently applied above regulatory safety enhancements in transport of spent nuclear fuel, this guidance did not adversely effect the Program. As the Program draws closer to its termination date, an increased number of requests for program extension are received. Currently, there are no plans to extend the policy beyond its current expiration date; therefore, eligible reactor operators interested in participating in this program are strongly encouraged to evaluate their inventory and plan for future shipments as soon as possible. (author)

  2. Modeling and Experimental Studies of Mercury Oxidation and Adsorption in a Fixed-Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buitrago, Paula A.; Morrill, Mike; Lighty, JoAnn S.; Silcox, Geoffrey D.

    2009-06-15

    This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO2, and halogen species were introduced through the burner to produce a radical pool representative of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO2, and NO2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and

  3. Analysis of a gas-liquid film plasma reactor for organic compound oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Hsieh, Kevin [Department of Chemical and Biomedical Engineering, Florida State University, Tallahassee, FL 32310 (United States); Wang, Huijuan [School of Environmental and Safety Engineering, Jiangsu University, Zhenjiang 212013 (China); Locke, Bruce R., E-mail: blocke@fsu.edu [Department of Chemical and Biomedical Engineering, Florida State University, Tallahassee, FL 32310 (United States)

    2016-11-05

    Highlights: • Non-homogeneous filamentary plasma discharge formed along gas-liquid interface. • Hydrogen peroxide formed near interface favored over organic oxidation from liquid. • Post-plasma Fenton reactions lead to complete utilization of hydrogen peroxide. - Abstract: A pulsed electrical discharge plasma formed in a tubular reactor with flowing argon carrier gas and a liquid water film was analyzed using methylene blue as a liquid phase hydroxyl radical scavenger and simultaneous measurements of hydrogen peroxide formation. The effects of liquid flow rate, liquid conductivity, concentration of dye, and the addition of ferrous ion on dye decoloration and degradation were determined. Higher liquid flow rates and concentrations of dye resulted in less decoloration percentages and hydrogen peroxide formation due to initial liquid conductivity effects and lower residence times in the reactor. The highest decoloration energy yield of dye found in these studies was 5.2 g/kWh when using the higher liquid flow rate and adding the catalyst. The non-homogeneous nature of the plasma discharge favors the production of hydrogen peroxide in the plasma-liquid interface over the chemical oxidation of the organic in the bulk liquid phase and post-plasma reactions with the Fenton catalyst lead to complete utilization of the plasma-formed hydrogen peroxide.

  4. Multi-objective optimization of oxidative desulfurization in a sono-photochemical airlift reactor.

    Science.gov (United States)

    Behin, Jamshid; Farhadian, Negin

    2017-09-01

    Response surface methodology (RSM) was employed to optimize ultrasound/ultraviolet-assisted oxidative desulfurization in an airlift reactor. Ultrasonic waves were incorporated in a novel-geometry reactor to investigate the synergistic effects of sono-chemistry and enhanced gas-liquid mass transfer. Non-hydrotreated kerosene containing sulfur and aromatic compounds was chosen as a case study. Experimental runs were conducted based on a face-centered central composite design and analyzed using RSM. The effects of two categorical factors, i.e., ultrasound and ultraviolet irradiation and two numerical factors, i.e., superficial gas velocity and oxidation time were investigated on two responses, i.e., desulfurization and de-aromatization yields. Two-factor interaction (2FI) polynomial model was developed for the responses and the desirability function associate with overlay graphs was applied to find optimum conditions. The results showed enhancement in desulfurization ability corresponds to more reduction in aromatic content of kerosene in each combination. Based on desirability approach and certain criteria considered for desulfurization/de-aromatization, the optimal desulfurization and de-aromatization yields of 91.7% and 48% were obtained in US/UV/O 3 /H 2 O 2 combination, respectively. Copyright © 2017 Elsevier B.V. All rights reserved.

  5. Continuous Polyol Synthesis of Metal and Metal Oxide Nanoparticles Using a Segmented Flow Tubular Reactor (SFTR

    Directory of Open Access Journals (Sweden)

    Andrea Testino

    2015-06-01

    Full Text Available Over the last years a new type of tubular plug flow reactor, the segmented flow tubular reactor (SFTR, has proven its versatility and robustness through the water-based synthesis of precipitates as varied as CaCO3, BaTiO3, Mn(1−xNixC2O4·2H2O, YBa oxalates, copper oxalate, ZnS, ZnO, iron oxides, and TiO2 produced with a high powder quality (phase composition, particle size, and shape and high reproducibility. The SFTR has been developed to overcome the classical problems of powder production scale-up from batch processes, which are mainly linked with mass and heat transfer. Recently, the SFTR concept has been further developed and applied for the synthesis of metals, metal oxides, and salts in form of nano- or micro-particles in organic solvents. This has been done by increasing the working temperature and modifying the particle carrying solvent. In this paper we summarize the experimental results for four materials prepared according to the polyol synthesis route combined with the SFTR. CeO2, Ni, Ag, and Ca3(PO42 nanoparticles (NPs can be obtained with a production rate of about 1–10 g per h. The production was carried out for several hours with constant product quality. These findings further corroborate the reliability and versatility of the SFTR for high throughput powder production.

  6. Destruction of an industrial wastewater by supercritical water oxidation in a transpiring wall reactor

    International Nuclear Information System (INIS)

    Bermejo, M.D.; Cocero, M.J.

    2006-01-01

    The supercritical water oxidation (SCWO) is a technology that takes advantage of the special properties of water in the surroundings of critical point of water to completely oxidize wastes in residence times lower than 1 min. The problems caused by the harsh operational conditions of the SCWO process are being solved by new reactor designs, such as the transpiring wall reactor (TWR). In this work, the operational parameters of a TWR have been studied for the treatment of an industrial wastewater. As a result, the process has been optimized for a feed flow of 16 kg/h with feed inlet temperatures higher than 300 deg. C and transpiring flow relation (R) between 0.2 and 0.6 working with an 8% (w/w) isopropanol (IPA) as a fuel. The experimental data and a mathematical model have been applied for the destruction of an industrial waste containing acetic acid and crotonaldehyde as main compounds. As the model predicted, removal efficiencies higher than 99.9% were obtained, resulting in effluents with 2 ppm total organic carbon (TOC) at feed flow of 16 kg/h, 320 deg. C of feed temperature and R = 0.32. An effluent TOC of 35 ppm under conditions feed flow of 18 kg/h, feed inlet temperatures of 290 deg. C, reaction temperatures of 570 deg. C and R = 0.6

  7. The characterization of a neutron radiography Triga reactor for NAA of chlorine in an iron oxide matrix

    International Nuclear Information System (INIS)

    Glagolenko, I.; Carney, K.; Difelici, R.; Maddison, D.; Sayer, M.; Hart, P.; Ross, J.; Kahn, S.; Swanson, R.

    2000-01-01

    An irradiation position in the 250 kW Triga reactor was characterized for instrumental neutron activation analysis of chlorine in an iron oxide matrix. Factors that affect the accuracy of the determination include variations in the reactor neutron spectrum and flux as a function of spatial position and the presence of chlorine impurities. Gold wire and foils were used to determine the neutron flux and cadmium ratio as a function of height in an air-filled irradiation tube. (author)

  8. Flow reactor studies of non-equilibrium plasma-assisted oxidation of n-alkanes.

    Science.gov (United States)

    Tsolas, Nicholas; Lee, Jong Guen; Yetter, Richard A

    2015-08-13

    The oxidation of n-alkanes (C1-C7) has been studied with and without the effects of a nanosecond, non-equilibrium plasma discharge at 1 atm pressure from 420 to 1250 K. Experiments have been performed under nearly isothermal conditions in a flow reactor, where reactive mixtures are diluted in Ar to minimize temperature changes from chemical reactions. Sample extraction performed at the exit of the reactor captures product and intermediate species and stores them in a multi-position valve for subsequent identification and quantification using gas chromatography. By fixing the flow rate in the reactor and varying the temperature, reactivity maps for the oxidation of fuels are achieved. Considering all the fuels studied, fuel consumption under the effects of the plasma is shown to have been enhanced significantly, particularly for the low-temperature regime (T<800 K). In fact, multiple transitions in the rates of fuel consumption are observed depending on fuel with the emergence of a negative-temperature-coefficient regime. For all fuels, the temperature for the transition into the high-temperature chemistry is lowered as a consequence of the plasma being able to increase the rate of fuel consumption. Using a phenomenological interpretation of the intermediate species formed, it can be shown that the active particles produced from the plasma enhance alkyl radical formation at all temperatures and enable low-temperature chain branching for fuels C3 and greater. The significance of this result demonstrates that the plasma provides an opportunity for low-temperature chain branching to occur at reduced pressures, which is typically observed at elevated pressures in thermal induced systems. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

  9. Status of University of Cincinnati reactor-site nuclear engineering graduate programs

    International Nuclear Information System (INIS)

    Anno, J.N.; Christenson, J.M.; Eckart, L.E.

    1993-01-01

    The University of Cincinnati (UC) nuclear engineering program faculty has now had 12 yr of experience in delivering reactor-site educational programs to nuclear power plant technical personnel. Currently, with the sponsorship of the Toledo-Edison Company (TED), we are conducting a multiyear on-site graduate program with more than 30 participants at the Davis-Besse nuclear power plant. The program enables TED employees with the proper academic background to earn a master of science (MS) degree in nuclear engineering (mechanical engineering option). This paper presents a brief history of tile evolution of UC reactor-site educational programs together with a description of the progress of the current program

  10. Foreign research reactor uranium supply program: The Y-12 national security complex process

    International Nuclear Information System (INIS)

    Nelson, T.; Eddy, B.G.

    2010-01-01

    The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the HEU Disposition Program, the Reduced Enrichment Research and Test Reactors (RERTR) Program, and the United States FRR Spent Nuclear Fuel (SNF) Acceptance Program. The Y-12 National Nuclear Security Administration (NNSA) Y-12 Site Office maintains the prime contracts with foreign governments for the supply of Low-Enriched Uranium (LEU) for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. This program supports the important U.S. government and nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to LEU fuel under the guidance of the NNSA RERTR Program. In conjunction with the FRR SNF Acceptance Program which supports the global nonproliferation efforts to disposition U.S.-origin HEU, the Y-12 FRR Uranium Supply Program can provide the LEU for the replacement fuel fabrication. In addition to feedstock for fuel fabrication, Y-12 supplies LEU for target fabrication for medical isotope production. The Y-12 process uses supply forecasting tools, production improvements and efficient delivery preparations to successfully support the global research reactor community

  11. p-Nitrophenol degradation by Fenton's oxidation in a bubble column reactor.

    Science.gov (United States)

    Rodrigues, Carmen S D; Borges, Ricardo A C; Lima, Vanessa N; Madeira, Luis M

    2018-01-15

    This paper reports on a study of the oxidation of p-nitrophenol (PNP) in a bubble column reactor (BCR). The use of the air stream aimed to provide perfect mixing in the liquid phase, which was successfully achieved and checked experimentally; there were no concentration gradients along the column, even at the lowest air flow rate used (Q = 1 mL/min at room temperature and atmospheric pressure). The effect of the operating variables was assessed, and a total reduction of PNP was reached, as well as mineralization of 49.2%, oxidant consumption of 90.3%, and with an efficiency of use - η H2O2 - of 0.09 mg C/mg H 2 O 2 , under the best operating conditions found - Q = 1 mL/min, [H 2 O 2 ] = 1.6 g/L, [Fe 2+ ] = 80 mg/L, pH = 3.0 and T = 22-24 °C - (after 120 min of reaction). Following this, various strategies were developed for improving the mineralization rate; it was found that the addition of H 2 O 2 every 5 min and readjusting the pH after 30 min of reaction allow the attainment of a much higher TOC removal (75.1%) and efficiency of oxidant use (η H2O2  = 0.17 mg C/mg H 2 O 2 ) with less oxidant. A reaction mechanism was proposed, based on intermediates identified that include p-nitrocatechol - PNC, p-benzoquinone - PB, hydroquinone - HQ - and carboxylic acids (oxalic, maleic and fumaric). Since the performance achieved in the BCR was good, and very similar to that obtained in a conventional batch reactor, it was possible to verify the efficacy of carrying out the Fenton process in this reactor configuration, which in our future work will focus on the treatability of industrial effluents. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Oxidation flow reactors (OFRs): overview of recent field and modeling studies

    Science.gov (United States)

    Jimenez, Jose-Luis; Palm, Brett B.; Peng, Zhe; Hu, Weiwei; Ortega, Amber M.; Li, Rui; Campuzano-Jost, Pedro; Day, Douglas A.; Stark, Harald; Brune, William H.; de Gouw, Joost; Schroder, Jason

    2016-04-01

    Oxidation flow reactors (OFRs) are popular tools for studying SOA formation and aging in both laboratory and field experiments. In an OFR, the concentration of an oxidant (OH, O3, or NO3) can be increased, leading to hours-months of equivalent atmospheric oxidation during the several-minute OFR residence time. Using gas- and particle-phase measurements from several recent field campaigns, we demonstrate SOA formation after oxidation of ambient air in an OFR. Typically, more SOA formation is observed from nighttime air than daytime air. This indicates that the concentration of SOA-forming gases in ambient air is relatively higher at night. Measured ambient VOCs are not able to explain the magnitude of SOA formation in the OFR, suggesting that typically unmeasured S/IVOCs (possibly VOC oxidation products or direct emissions) play a substantial intermediary role in ambient SOA formation. We also present highlights from recent OFR oxidant chemistry modeling studies. HOx, Ox, and photolysis chemistry was modeled for two common OH production methods (utilizing 185+254 nm UV light, or 254 nm only). OH exposure (OHexp) can be estimated within a factor of ~2 using model-derived equations, and can be verified in situ using VOC decay measurements. OHexp is strongly dependent on external OH reactivity, which may cause significant OH suppression in some circumstances (e.g., lab/source studies with high precursor concentrations). UV light photolysis and reaction with oxygen atoms are typically not major reaction pathways. Modeling the fate of condensable low-volatility organic gases (LVOCs) formed in an OFR suggests that LVOC fate is dependent on particle condensational sink. E.g., for the range of particle condensational sink at a remote pine forest, anywhere from 20-80% of produced LVOCs were predicted to condense onto aerosols for an OHexp of ~1 day, with the remainder lost to OFR or sampling line walls. Similar to large chamber wall loss corrections, a correction is needed

  13. Development, application and also modern condition of the calculated program Imitator of a reactor

    International Nuclear Information System (INIS)

    Aver'yanova, S.P.; Kovel', A.I.; Mamichev, V.V.; Filimonov, P.E.

    2008-01-01

    Features of the calculated program Imitator of a reactor (IR) for WWER-1000 operation simulation are discussed. It is noted that IR application at NPP provides for the project program (BIPR-7) on-line working. This offers a new means, on the one hand, for the efficient prediction and information support of operator, on the other hand, for the verification and development of calculated scheme and neutron-physical model of the WWER-1000 projection program [ru

  14. Study on the seismic verification test program on the experimental multi-purpose high-temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Taketani, K.; Aochi, T.; Yasuno, T.; Ikushima, T.; Shiraki, K.; Honma, T.; Kawamura, N.

    1978-01-01

    The paper describes a program of experimental research necessary for qualitative and quantitative determination of vibration characteristics and aseismic safety on structure of reactor core in the multipurpose high temperature gas-cooled experimental reactor (VHTR Experimental Reactor) by the Japan Atomic Energy Research Institute

  15. University Reactor Instrumentation grant program. Final report, September 7, 1990--August 31, 1995

    International Nuclear Information System (INIS)

    Talnagi, J.W.

    1998-01-01

    The Ohio State University Nuclear Reactor Laboratory (OSU NRL) participated in the Department of Energy (DOE) grant program commonly denoted as the University Reactor Instrumentation (URI) program from the period September 1990 through August 1995, after which funding was terminated on a programmatic basis by DOE. This program provided funding support for acquisition of capital equipment targeted for facility upgrades and improvements, including modernizing reactor systems and instrumentation, improvements in research and instructional capabilities, and infrastructure enhancements. The staff of the OSU NRL submitted five grant applications during this period, all of which were funded either partially or in their entirety. This report will provide an overview of the activities carried out under these grants and assess their impact on the OSU NRL facilities

  16. Optimizing The Efficiency of a Dielectric Barrier Discharge Reactor for Removal of Nitric Oxides in Gas Phase

    International Nuclear Information System (INIS)

    Siti Aiasah Hashim; Wong, C.S.; Abas, M.R.

    2016-01-01

    A dielectric barrier discharge (DBD) reactor was built and used to remove nitric oxides in gas phase. In the preliminary work, it was found that the DBD reactor can used for direct processing of contaminated air stream. It was observed that if the applied energy is sufficiently high, reduction can overcome the oxidation process. The other characteristics that can affect the efficiency of the reactor are the processing flow rate, number of DBD tubes used and how the tubes are connected. The composition of the feed gas also plays important role. To improve the efficiency, more tubes were added and configured in combination of serial and parallel connections to achieve the best result. The reactor was found to be most efficient when using 6 tubes configured to have 2 sets of 3 tubes in series connected in parallel. The maximum flow rate that can be treated is 5 scfh. When operated with the optimum input voltage of 32 kV, the reactor can remove up to 80 % nitric oxide in the reduction mode. This means that the energy is sufficiently high to sustain the reduction mode and prevent further oxidation. (author)

  17. Evaluation of tubular reactor designs for supercritical water oxidation of U.S. Department of Energy mixed waste

    International Nuclear Information System (INIS)

    Barnes, C.M.

    1994-12-01

    Supercritical water oxidation (SCWO) is an emerging technology for industrial waste treatment and is being developed for treatment of the US Department of Energy (DOE) mixed hazardous and radioactive wastes. In the SCWO process, wastes containing organic material are oxidized in the presence of water at conditions of temperature and pressure above the critical point of water, 374 C and 22.1 MPa. DOE mixed wastes consist of a broad spectrum of liquids, sludges, and solids containing a wide variety of organic components plus inorganic components including radionuclides. This report is a review and evaluation of tubular reactor designs for supercritical water oxidation of US Department of Energy mixed waste. Tubular reactors are evaluated against requirements for treatment of US Department of Energy mixed waste. Requirements that play major roles in the evaluation include achieving acceptable corrosion, deposition, and heat removal rates. A general evaluation is made of tubular reactors and specific reactors are discussed. Based on the evaluations, recommendations are made regarding continued development of supercritical water oxidation reactors for US Department of Energy mixed waste

  18. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1982-August 1983

    International Nuclear Information System (INIS)

    Brenizer, J.S.; Benneche, P.E.

    1984-03-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1982 was over 5500 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1982 through August 1983

  19. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1983-August 1984

    International Nuclear Information System (INIS)

    Mulder, R.U.; Benneche, P.E.

    1984-11-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1983 was over 6000 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1983 through August 1984

  20. Reactor Physics Experiments by Korean Under-Graduate Students in Kyoto University Critical Assembly Program (KUGSiKUCA Program)

    International Nuclear Information System (INIS)

    Pyeon, Cheol Ho; Misawa, Tsuyoshi; Unesaki, Hironobu; Ichihara, Chihiro; Shiroya, Seiji; Whang, Joo Ho; Kim, Myung Hyun

    2006-01-01

    The Reactor Laboratory Course for Korean Under-Graduate Students in Kyoto University Critical Assembly (KUGSiKUCA) program has been launched from 2003, as one of international collaboration programs of Kyoto University Research Reactor Institute (KURRI). This program was suggested by Department of Nuclear Engineering, College of Advanced Technology, Kyunghee University (KHU), and was adopted by Ministry of Science and Technology of Korean Government as one of among Nuclear Human Resources Education and Training Programs. On the basis of her suggestion for KURRI, memorandum for academic corporation and exchange between KHU and KURRI was concluded on July 2003. The program has been based on the background that it is extremely difficult for any single university in Korea to have her own research or training reactor. Up to this 2006, total number of 61 Korean under-graduate school students, who have majored in nuclear engineering of Kyunghee University, Hanyang University, Seoul National University, Korea Advanced Institute of Science and Technology, Chosun University and Cheju National University in all over the Korea, has taken part in this program. In all the period, two professors and one teaching assistant on the Korean side led the students and helped their successful experiments, reports and discussions. Due to their effort, the program has succeeded in giving an effective and unique course, taking advantage of their collaboration

  1. 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.

  2. Immediate relation of ING to fast breeder reactor programs

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W B

    1969-07-01

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  3. Immediate relation of ING to fast breeder reactor programs

    International Nuclear Information System (INIS)

    Lewis, W.B.

    1969-01-01

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  4. The Thermos program for nuclear reactors specialized in district heating

    International Nuclear Information System (INIS)

    Lerouge, B.

    1976-01-01

    Many studies have been made in France on the use of nuclear heat for district heating. After a brief account of the problems raised by the use of thermal waste from big nuclear power stations, the quantitative and qualitative needs of heating networks are analyzed and the Thermos project described. This is a very robust reactor of the pool type, with an output of 100MW, supplying low-pressure water at 100 deg C. The advantages from the aspects of safety and economy are described, and the present state of the project and its possible developments summarized [fr

  5. New approaches in the USA for high-temperature gas-cooled reactors. Gas-Cooled Reactor Programs

    International Nuclear Information System (INIS)

    Kasten, P.R.; Neylan, A.J.; Penfield, S.R. Jr.

    1985-08-01

    Several concepts are being evaluated in the US HTR Program to explore designs which might improve the commercial viability of nuclear power. The general approach is to reduce the reactor power and increase the ability to use inherent features for removing heat following extreme accidents. The unit size and design of these concepts are constrained so that extreme accidents do not result in significant release of radioactivity from the reactor plant. Through the greater reliance on inherent safety features in small HTRs, it should be possible to minimize the amount of nuclear grade components required in the balance-of-plant, which could lead to an economic system. Four HTR concepts are presently being evaluated within the US Program, and these concepts are briefly summarized. A modular HTR using a steel pressure vessel, which is very similar to one of the four HTR concepts being evaluated within the US National program, is presented as an example of a specific concept to illustrate the features and performance of HTRs having a high degree of inherent safety

  6. Fast reactor test facilities in the US safety program

    International Nuclear Information System (INIS)

    Avery, R.; Dickerman, C.E.; Lennox, D.H.; Rose, D.

    1979-01-01

    The needs for safety information derivable from in-pile programs are reviewed, and the correlation made with existing and planned capability. In view of the current status of the U.S. breeder program, emphasis is given in the review to the impact of different fast breeder options on the required program and facilities. It is concluded that facility needs are somewhat independent of specific fast breeder concept, even though the relative emphasis on the various safety issues will differ. 8 refs

  7. The development of quality assurance program in Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Rosli Darmawan; Mohd Rizal Mamat; Mohamad Zaid Mohamad; Mohd Ridzuan Abdul Mutalib

    2007-01-01

    One of the trivial issues in the operation of Nuclear Reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing has always bring about general public fear on anything related to nuclear. IAEA has always emphasized on the assurance of nuclear safety for all nuclear installations and activities. According to the IAEA safety guides, all research reactors are required to implement quality assurance programs to ensure the conduct of operations are in accordance with the safety standards required. This paper discusses the activities carried out toward the establishment of Quality Assurance Program for Reaktor TRIGA PUSPATI (RTP). (Author)

  8. University Reactor Sharing Program. Period covered: September 1, 1981-August 31, 1982

    International Nuclear Information System (INIS)

    Hajek, B.K.; Myser, R.D.; Miller, D.W.

    1982-12-01

    During the period from September 1, 1981 to August 31, 1982, the Ohio State University Nuclear Reactor Laboratory participated in the Reactor Sharing Program by providing services to eight colleges and universities. A laboratory on Neutron Activation Analysis was developed for students in the program. A summary of services provided and a copy of the laboratory procedure are attached. Services provided in the last funded period were in three major areas. These were neutron activation analysis, nuclear engineering labs, and introductions to nuclear research. One group also performed radiation surveys and produced isotopes for calibration of their own analytical equipment

  9. Review of Savannah River Site K Reactor inservice inspection and testing restart program

    International Nuclear Information System (INIS)

    Anderson, M.T.; Hartley, R.S.; Kido, C.

    1992-09-01

    Inservice inspection (ISI) and inservice testing (IST) programs are used at commercial nuclear power plants to monitor the pressure boundary integrity and operability of components in important safety-related systems. The Department of Energy (DOE) - Office of Defense Programs (DP) operates a Category A (> 20 MW thermal) production reactor at the Savannah River Site (SRS). This report represents an evaluation of the ISI and IST practices proposed for restart of SRS K Reactor as compared, where applicable, to current ISI/IST activities of commercial nuclear power facilities

  10. Development programs on decommissioning technology for reactors and fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Fujiki, K.

    1992-01-01

    The Science and Technology Agency (STA) of Japan is promoting technology development for decommissioning of nuclear facilities by entrusting various research programs to concerned research organisations: JAERI, PNC and RANDEC, including first full scale reactor decommissioning of JPDR. According to the results of these programs, significant improvement on dismantling techniques, decontamination, measurement etc. has been achieved. Further development of advanced decommissioning technology has been started in order to achieve reduction of duration of decommissioning work and occupational exposures in consideration of the decommissioning of reactors and fuel cycle facilities. (author) 5 refs.; 7 figs.; 1 tab

  11. Research program of the high temperature engineering test reactor for upgrading the HTGR technology

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

    1997-07-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

  12. United States Department of Energy breeder reactor staff training domestic program

    International Nuclear Information System (INIS)

    1984-01-01

    Two US DOE projects in the Pacific Northwest offer unique on-the-scene training opportunities at sodium-cooled fast-reactor plants: the Fast Flux Test Facility (FFTF) near Richland, Washington, which has operated successfully in a wide range of irradiation test programs since 1980; and the Experimental Breeder Reactor II (EBR-II) near Idaho Falls, Idaho, which has been in operation for approximately 20 years. Training programs have been especially designed to take advantage of this plant experience. Available courses are described

  13. Reacting flow simulations of supercritical water oxidation of PCB-contaminated transformer oil in a pilot plant reactor

    Directory of Open Access Journals (Sweden)

    V. Marulanda

    2011-06-01

    Full Text Available The scale-up of a supercritical water oxidation process, based on recent advancements in kinetic aspects, reactor configuration and optimal operational conditions, depends on the research and development of simulation tools, which allow the designer not only to understand the complex multiphysics phenomena that describe the system, but also to optimize the operational parameters to attain the best profit for the process and guarantee its safe operation. Accordingly, this paper reports a multiphysics simulation with the CFD software Comsol Multiphysics 3.3 of a pilot plant reactor for the supercritical water oxidation of a heavily PCB-contaminated mineral transformer oil. The proposed model was based on available information for the kinetic aspects of the complex mixture and the optimal operational conditions obtained in a lab-scale continuous supercritical water oxidation unit. The pilot plant simulation results indicate that it is not feasible to scale-up directly the optimal operational conditions obtained in the isothermal lab-scale experiments, due to the excess heat released by the exothermic oxidation reactions that result in outlet temperatures higher than 600°C, even at reactor inlet temperatures as low as 400°C. Consequently, different alternatives such as decreasing organic flowrates or a new reactor set-up with multiple oxidant injections should be considered to guarantee a safe operation.

  14. Oxidation of organics in water in microfluidic electrochemical reactors: Theoretical model and experiments

    International Nuclear Information System (INIS)

    Scialdone, Onofrio; Guarisco, Chiara; Galia, Alessandro

    2011-01-01

    The electrochemical oxidation of organics in water performed in micro reactors on boron doped diamond (BDD) anode was investigated both theoretically and experimentally in order to find the influence of various operative parameters on the conversion and the current efficiency CE of the process. The electrochemical oxidation of formic acid (FA) was selected as a model case. High conversions for a single passage of the electrolytic solution inside the cell were obtained by operating with proper residence times and low distances between cathode and anode. The effect of initial concentration, flow rate and current density was investigated in detail. Theoretical predictions were in very good agreement with experimental results for both mass transfer control, oxidation reaction control and mixed kinetic regimes in spite of the fact that no adjustable parameters was used. Mass transfer process was successfully modelled by considering for simplicity a constant Sh number (e.g., a constant mass transfer coefficient k m ) for a process performed with no high values of the current intensity to minimize the effect of the gas bubbling on the flowdynamic pattern. For mixed kinetic regimes, two different modelling approaches were used. In the first one, the oxidation of organics at BDD was assumed to be mass transfer controlled and to occur with an intrinsic 100% CE when applied current density is higher than the limiting current density. In the second case, the CE of the process was modelled assuming that the competition between organic and water oxidation depends only on the electrodic material and on the nature and the concentration of the organic. In the latter case a better agreement between experimental data and theoretical predictions was observed.

  15. Progress of the RERTR [Reduced Enrichment Research and Test Reactor] Program in 1989

    International Nuclear Information System (INIS)

    Travelli, A.

    1989-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs

  16. International topical meeting on research reactor fuel management (RRFM) - United States foreign research reactor (FRR) spent nuclear fuel (SNF) acceptance program: 2010 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.; Niehus, M.T.; Landers, C.

    2010-01-01

    The Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by the United States Department of Energy (DOE), in consultation with the Department of State (DOS) in May 1996, scheduled to expire May 12, 2016, to return research reactor fuel until May 12, 2019 to the U.S. is in its fourteenth year. This paper provides a brief update on the program, part of the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of U.S.-origin nuclear materials, which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible research reactors who have not yet participated in the program. (author)

  17. Optimization of fuel management and control poison of a nuclear power reactor by dynamic programming

    International Nuclear Information System (INIS)

    Lima, C.A.R. de.

    1977-01-01

    The distribution of fuel and control poison in a nuclear reactor was optimized by the method of Dynamic Programming. A 620 M We Pressurized Water Reactor similar to Angra-1 was studied. The reactor operation was simulated in a IBM-1130 computer. Two fuel shuffling schemes and three poison management schemes were simultaneously employed in the reactor divided into three regions of equal volume and two consecutive stages were studied in order to determine the influence of poison management on the optimum fuel management policy. When uniform poisoning on all the three regions was permitted the traditional out-in fuel management policy proved to be more economic. On introducing simultaneous poison management, the optimum fuel management sequence was found to be different. The results obtained indicate a stronger interaction between the fuel management and the poison management than anticipated in previous works. (author)

  18. Low nitrous oxide production through nitrifier-denitrification in intermittent-feed high-rate nitritation reactors

    DEFF Research Database (Denmark)

    Su, Qingxian; Ma, Chun; Domingo-Felez, Carlos

    2017-01-01

    Nitrous oxide (N2O) production from autotrophic nitrogen conversion processes, especially nitritation systems, can be significant, requires understanding and calls for mitigation. In this study, the rates and pathways of N2O production were quantified in two lab-scale sequencing batch reactors...... to maintain high nitritation efficiency and high nitritation rates at 20-26 °C over a period of ∼300 days. Even at the high nitritation efficiencies, net N2O production was low (∼2% of the oxidized ammonium). Net N2O production rates transiently increased with a rise in pH after each feeding, suggesting...... operated with intermittent feeding and demonstrating long-term and high-rate nitritation. The resulting reactor biomass was highly enriched in ammonia-oxidizing bacteria, and converted ∼93 ± 14% of the oxidized ammonium to nitrite. The low DO set-point combined with intermittent feeding was sufficient...

  19. Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Wood, J.C.

    1987-06-01

    Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

  20. Jet-stirred reactor oxidation of alkane-rich FACE gasoline fuels

    KAUST Repository

    Chen, Bingjie

    2016-06-23

    Understanding species evolution upon gasoline fuel oxidation can aid in mitigating harmful emissions and improving combustion efficiency. Experimentally measured speciation profiles are also important targets for surrogate fuel kinetic models. This work presents the low- and high-temperature oxidation of two alkane-rich FACE gasolines (A and C, Fuels for Advanced Combustion Engines) in a jet-stirred reactor at 10. bar and equivalence ratios from 0.5 to 2 by probe sampling combined with gas chromatography and Fourier Transformed Infrared Spectrometry analysis. Detailed speciation profiles as a function of temperature are presented and compared to understand the combustion chemistry of these two real fuels. Simulations were conducted using three surrogates (i.e., FGA2, FGC2, and FRF 84), which have similar physical and chemical properties as the two gasolines. The experimental results reveal that the reactivity and major product distributions of these two alkane-rich FACE fuels are very similar, indicating that they have similar global reactivity despite their different compositions. The simulation results using all the surrogates capture the two-stage oxidation behavior of the two FACE gasolines, but the extent of low temperature reactivity is over-predicted. The simulations were analyzed, with a focus on the n-heptane and n-butane sub-mechanisms, to help direct the future model development and surrogate fuel formulation strategies.

  1. Jet-stirred reactor oxidation of alkane-rich FACE gasoline fuels

    KAUST Repository

    Chen, Bingjie; Togbé , Casimir; Wang, Zhandong; Dagaut, Philippe; Sarathy, Mani

    2016-01-01

    Understanding species evolution upon gasoline fuel oxidation can aid in mitigating harmful emissions and improving combustion efficiency. Experimentally measured speciation profiles are also important targets for surrogate fuel kinetic models. This work presents the low- and high-temperature oxidation of two alkane-rich FACE gasolines (A and C, Fuels for Advanced Combustion Engines) in a jet-stirred reactor at 10. bar and equivalence ratios from 0.5 to 2 by probe sampling combined with gas chromatography and Fourier Transformed Infrared Spectrometry analysis. Detailed speciation profiles as a function of temperature are presented and compared to understand the combustion chemistry of these two real fuels. Simulations were conducted using three surrogates (i.e., FGA2, FGC2, and FRF 84), which have similar physical and chemical properties as the two gasolines. The experimental results reveal that the reactivity and major product distributions of these two alkane-rich FACE fuels are very similar, indicating that they have similar global reactivity despite their different compositions. The simulation results using all the surrogates capture the two-stage oxidation behavior of the two FACE gasolines, but the extent of low temperature reactivity is over-predicted. The simulations were analyzed, with a focus on the n-heptane and n-butane sub-mechanisms, to help direct the future model development and surrogate fuel formulation strategies.

  2. Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Mejia C, M. A.

    2016-01-01

    The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

  3. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    International Nuclear Information System (INIS)

    McCarthy, K.A.; Williams, D.L.; Reister, R.

    2012-01-01

    The US Department of Energy Light Water Reactor Sustainability (LWRS) Program is focused on enabling the long-term operation of US commercial power plants. Decisions on life extension will be made by commercial power plant owners - the information provided by the research and development activities in the LWRS Program will reduce the uncertainty (and therefore the risk) associated with making those decisions. The LWRS Program encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper provides an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables. (author)

  4. Development status and potential program for development of proliferation-resistant molten-salt reactors

    International Nuclear Information System (INIS)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review

  5. Development status and potential program for development of proliferation-resistant molten-salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review.

  6. Russian research reactor fuel return program starts shipping fuel to Russia

    International Nuclear Information System (INIS)

    Dedik, T.; Bolshinsky, I.; Krass, A.

    2003-01-01

    For almost four years the United States (U.S), the Russian Federation (R.F.), and the International Atomic Energy Agency (IAEA) have been discussing an initiative to return Soviet/Russian-origin research reactor fuel to the Russian Federation. In a series of bilateral and trilateral meetings in Vienna and Moscow, considerable progress has been made toward defining the Russian Research Reactor Fuel Return Program as well as obtaining the necessary technical data to facilitate the return. More than 20 research reactors in 17 countries that have Soviet- or Russian-supplied fuel have identified. Most of these reactors have stocks of both fresh and irradiated HEU fuel that must be carefully stored and managed for many years to come. On September 21, 2003 the Russian Research Reactor Fuel Return program shipped 14 kg of fresh Russian-origin HEU fuel from Romania to the nuclear fuel fabrication facility in Russia, which represented the beginning of the practical implementation of the program. (author)

  7. The irradiation test program for transmutation in the French Phenix fast reactor

    International Nuclear Information System (INIS)

    Guidez, J.; Chaucheprat, P.; Fontaine, B.; Brunon, E.

    2004-01-01

    Put on commercial operation in July 1974, the French fast reactor Phenix reached a 100 000 hours operation time in september 2003. When the French law relative to long lived radioactive waste management was promulgated on December 1991, priority was given to Phenix to be run as a research reactor and to carry on a wide irradiation program dedicated to study transmutation of minor actinides and long-lived fission products. After a major renovation program required to extend the reactor lifetime, Phenix power buildup took place in 2003. Experimental irradiations have been loaded in the core, involving components for heterogeneous and homogeneous transmutation modes, americium targets, technetium 99 metal pins and isolated isotopes for integral cross-sections measurements. Associated post- irradiated examination programs are already underway or planned. With new experiments to be loaded in the core in 2006 the Phenix reactor remains to be a powerful tool providing an important experimental data on fast reactors and on transmutation of minor actinides and long-lived fission products, as well as it will contribute to gain further experience in the framework of the GENERATION IV International Forum. (authors)

  8. Use of multiple on-campus reactors in education and training programs

    International Nuclear Information System (INIS)

    Schlapper, G.A.

    1989-01-01

    In its undergraduate and graduate programs in nuclear engineering and health physics, Texas A ampersand M University utilizes two reactors for the training and education of students. The 5-W AGN-201 nuclear training reactor has been in use since the late 1950s, while the 1-MW TRIGA Nuclear Science Center Reactor (NSCR) was first utilized in late 1961. Both facilities have been upgraded since initial criticality, the AGN power level being increased from the original 200-mW limit to its 5-W current level and the NSCR undergoing conversion from a 100-kW materials test reactor fueled deign to a 1-MW TRIGA-fueled facility. The AGN reactor is operated by the Department of Nuclear Engineering of the College of Engineering and is almost solely utilized in training and education programs. The NSCR facility is administered by the Texas Engineering Experiment Station and support research efforts of faculty and students of departments within and outside the university in addition to contributing to the education and training programs of the nuclear engineering department

  9. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  10. Current status of Westinghouse tubular solid oxide fuel cell program

    Energy Technology Data Exchange (ETDEWEB)

    Parker, W.G. [Westinghouse Science and Technology Center, Pittsburgh, PA (United States)

    1996-04-01

    In the last ten years the solid oxide fuel cell (SOFC) development program at Westinghouse has evolved from a focus on basic material science to the engineering of fully integrated electric power systems. Our endurance for this cell is 5 to 10 years. To date we have successfully operated at power for over six years. For power plants it is our goal to have operated before the end of this decade a MW class power plant. Progress toward these goals is described.

  11. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1981-August 1982

    International Nuclear Information System (INIS)

    Brenizer, J.S.; Benneche, P.E.

    1982-12-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1981 and nearly 5000 megawatt-hours) research reactor in the mid-Atlantic States. In addition, a second, small (50 watt) reactor is also available for use in educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1981 through August 1982

  12. A mathematical method for boiling water reactor control rod programming

    International Nuclear Information System (INIS)

    Tokumasu, S.; Hiranuma, H.; Ozawa, M.; Yokomi, M.

    1985-01-01

    A new mathematical programming method has been developed and utilized in OPROD, an existing computer code for automatic generation of control rod programs as an alternative inner-loop routine for the method of approximate programming. The new routine is constructed of a dual feasible direction algorithm, and consists essentially of two stages of iterative optimization procedures Optimization Procedures I and II. Both follow almost the same algorithm; Optimization Procedure I searches for feasible solutions and Optimization Procedure II optimizes the objective function. Optimization theory and computer simulations have demonstrated that the new routine could find optimum solutions, even if deteriorated initial control rod patterns were given

  13. Reactor operations, inspection and maintenance. PNGS Calibration Program

    International Nuclear Information System (INIS)

    Lopez, E.

    1997-01-01

    The PNGS Calibration Program is being implemented as a response to various concerns identified in recent PEER evaluations and AECB audits. Identified areas of concern were the approach to instrument calibration of Special Safety Systems (SSS). The implementation of a calibration program is a significant improvement in operating practices. A systematic and comprehensive approach to calibration of instrumentation will improve the quality of operation of the plant with a positive contribution to PNGS safety of operation and economic objectives. This paper describes the strategy to implement the proposed calibration program and describes its calibration data requirements. (DM)

  14. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  15. Fast reactor development program in France in 1999

    International Nuclear Information System (INIS)

    Leclere, Jacques; Teilland, C.; Astegiano, J.C.; Desreumaux, J.; Grenet, P.; Del Beccaro, R.; Vasile, A.

    2000-01-01

    The year 1999 was marked by important industrial reorganisations, especially concerning Framatome: the shareholding was modified, and at the end of the year the fusion of nuclear activities with those of Siemens was decided. The construction of the last N4 plants is completed, with no realisation expected in the near future; the decision on an EPR unit is not expected in the short term. The operation of the plants was satisfying, the availability is nevertheless lower than in 1998. Renovation, inspection and maintenance works were actively conducted at Phenix; decommissioning work was undertaken at Superphenix; the core unloading began in December. The R and D strategy was redefined; on fast reactors, work decreases on sodium and increases on other variants. International collaborations remain active. (author)

  16. Gas reactor international cooperative program. HTR-synfuel application assessment

    International Nuclear Information System (INIS)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H 2 +CO 2 ) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000 0 F steam is generated at the industrial user sites. The products of methanation (CH 4 + H 2 O) are piped back to the reformer at the central station HTR

  17. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  18. A spare-parts inventory program for TRIGA reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, T V; Ringle, J C; Johnson, A G [Oregon State University (United States)

    1974-07-01

    As is fairly common with new reactor facilities, we had a few spare parts on hand as part of our original purchase when the OSU TRIGA first went critical in March of 1967. Within a year or so, however, it became apparent that we should critically examine our spare parts inventory in order to avoid unnecessary or prolonged outages due to lack of a crucial piece of equipment. Many critical components (those which must be present and operable according to our license or technical specifications) were considered, and a priority list of acquiring these was established. This first list was drawn up in March, 1969, two years after initial criticality, and some key components were ordered. The availability of funds was the overriding restriction then and now. This spare-parts list is reviewed and new components purchased annually; the average amount spent has been about $2,000 per year. This inventory has proved invaluable more than once; without it, we would have had lengthy shutdowns awaiting the arrival of the needed component. The sobering thought, however, is that our spare-parts inventory is still not complete-far from it, in fact, because this would be prohibitively expensive. It is very difficult to guess with 100% accuracy just which component might need replacing, and your $10,000 inventory of spare parts is useless in that instance if it doesn't include the needed part. An idea worth considering is to either (a) encourage General Atomic, through the collective voice of all TRIGA owners, to maintain a rather complete inventory of replacement parts, or (b) maintain an owner's spare-parts pool, financed by contributions from all the facilities. If either of these pools was established, the needed part could reach any facility within the U.S. within a few days, minimizing reactor outage time. (author)

  19. A spare-parts inventory program for TRIGA reactors

    International Nuclear Information System (INIS)

    Anderson, T.V.; Ringle, J.C.; Johnson, A.G.

    1974-01-01

    As is fairly common with new reactor facilities, we had a few spare parts on hand as part of our original purchase when the OSU TRIGA first went critical in March of 1967. Within a year or so, however, it became apparent that we should critically examine our spare parts inventory in order to avoid unnecessary or prolonged outages due to lack of a crucial piece of equipment. Many critical components (those which must be present and operable according to our license or technical specifications) were considered, and a priority list of acquiring these was established. This first list was drawn up in March, 1969, two years after initial criticality, and some key components were ordered. The availability of funds was the overriding restriction then and now. This spare-parts list is reviewed and new components purchased annually; the average amount spent has been about $2,000 per year. This inventory has proved invaluable more than once; without it, we would have had lengthy shutdowns awaiting the arrival of the needed component. The sobering thought, however, is that our spare-parts inventory is still not complete-far from it, in fact, because this would be prohibitively expensive. It is very difficult to guess with 100% accuracy just which component might need replacing, and your $10,000 inventory of spare parts is useless in that instance if it doesn't include the needed part. An idea worth considering is to either (a) encourage General Atomic, through the collective voice of all TRIGA owners, to maintain a rather complete inventory of replacement parts, or (b) maintain an owner's spare-parts pool, financed by contributions from all the facilities. If either of these pools was established, the needed part could reach any facility within the U.S. within a few days, minimizing reactor outage time. (author)

  20. Fast reactor safety program. Progress report, January-March 1980

    International Nuclear Information System (INIS)

    1980-05-01

    The goal of the DOE LMFBR Safety Program is to provide a technology base fully responsive to safety considerations in the design, evaluation, licensing, and economic optimization of LMFBRs for electrical power generation. A strategy is presented that divides safety technology development into seven program elements, which have been used as the basis for the Work Breakdown Structure (WBS) for the Program. These elements include four lines of assurance (LOAs) involving core-related safety considerations, an element supporting non-core-related plant safety considerations, a safety R and D integration element, and an element for the development of test facilities and equipment to be used in Program experiments: LOA-1 (prevent accidents); LOA-2 (limit core damage); LOA-3 (maintain containment integrity); LOA-4 (attenuate radiological consequences); plant considerations; R and D integration; and facility development

  1. The Alternate Technology Program for Aluminum Research Reactor Spent Fuel

    International Nuclear Information System (INIS)

    Barlow, M.W.

    1998-01-01

    This paper describes the program for disposition of aluminum-based RRSNF, including the requirements for road-ready dry storage and repository disposal and the criteria to be considered in selecting among the alternative technologies

  2. High-temperature gas-cooled reactor safety-reliability program plan

    Energy Technology Data Exchange (ETDEWEB)

    1981-03-01

    The purpose of this document is to present a safety plan as part of an overall program plan for the design and development of the High Temperature Gas-Cooled Reactor (HTGR). This plan is intended to establish a logical framework for identifying the technology necessary to demonstrate that the requisite degree of public risk safety can be achieved economically. This plan provides a coherent system safety approach together with goals and success criterion as part of a unifying strategy for licensing a lead reactor plant in the near term. It is intended to provide guidance to program participants involved in producing a technology base for the HTGR that is fully responsive to safety consideration in the design, evaluation, licensing, public acceptance, and economic optimization of reactor systems.

  3. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

  4. Annual report in compliance with the reactor sharing program, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    Karam, R.A.

    1997-01-01

    This report contains information with regard to facilities utilization, descriptions (brief), personnel, organization, and programs of the Neely Nuclear Research Center (NNRC) at the Georgia Institute of Technology. The NNRC has two major facilities: the Georgia Tech Research Reactor and the Hot Cell Laboratory. This report of NNRC utilization is prepared in compliance with the contract requirements between the U.S. Department of Energy and the Georgia Institute of Technology. The NNRC is a participant in the University Reactor Sharing Program; as such, it makes available its 5 MW research reactor, its Co-60 irradiation facility and its activation analysis laboratory to large numbers of students and faculty from many universities and colleges

  5. Area Safety Program for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Rappe, G.M.

    1984-10-01

    Overall the Area Safety Program has proved to be a very successful operation. There is no doubt that a safety program organized through line management is the best way to involve all personnel. Naturally, when the program was first started, there was some criticism and a certain resistance on the part of a few individuals to fully participate. However, once the program was underway and it could be seen that it was working to everyone's advantage, this reluctance disappeared and a spirit of full cooperation is now enjoyed. It is very important that for this success to continue there must be a two way flow of information, both from the Area Safety Coordinators up through line management, and from senior management, with decisions and answers, back down through the management chain with the utmost dispatch. As with all programs, there is still room for improvement. This program has started a review cycle with a view to streamlining certain areas and possibly increasing its scope in others

  6. The role of a technology demonstration program for future reactors

    International Nuclear Information System (INIS)

    Viktorov, A.

    2011-01-01

    A comprehensive technology demonstration program is seen as an important component of the overall safety case, especially for a novel technology. The objective of such a program is defined as providing objective and auditable evidence that the technology will meet or exceed the relevant requirements. Various aspects of such a program are identified and then discussed in some details in this presentation. We will show how the need for such a program is anchored in fundamental safety principles. Attributes of the program, means of achieving its objective, roles of participants, as well as key steps are all elaborated. It will be argued that to prove a novel technology, the designer will have to combine several activities such as the use of operational experience, prototyping of the technology elements, conduct of experiments and tests under representative conditions, as well as modeling and analysis. Importance of availability of experimental facilities and qualified scientific and technical staff is emphasized. A solid technology demonstration program will facilitate and speed up regulatory evaluations of licensing applications. (author)

  7. The role of a technology demonstration program for future reactors

    Energy Technology Data Exchange (ETDEWEB)

    Viktorov, A. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2011-07-01

    A comprehensive technology demonstration program is seen as an important component of the overall safety case, especially for a novel technology. The objective of such a program is defined as providing objective and auditable evidence that the technology will meet or exceed the relevant requirements. Various aspects of such a program are identified and then discussed in some details in this presentation. We will show how the need for such a program is anchored in fundamental safety principles. Attributes of the program, means of achieving its objective, roles of participants, as well as key steps are all elaborated. It will be argued that to prove a novel technology, the designer will have to combine several activities such as the use of operational experience, prototyping of the technology elements, conduct of experiments and tests under representative conditions, as well as modeling and analysis. Importance of availability of experimental facilities and qualified scientific and technical staff is emphasized. A solid technology demonstration program will facilitate and speed up regulatory evaluations of licensing applications. (author)

  8. Spectrographic determination of impurities in ceramic materials for nuclear fusion reactors III. Analysis of magnesium oxide

    International Nuclear Information System (INIS)

    Melon, A. M.; Roca, M.; Rucandio, M. I.

    1992-01-01

    The determination of minor and trace elements in the magnesium oxide, considered as possible ceramic material in thermonuclear fusion reactors, has been studied. The concentration ranges are 0.1 - 0.3% for Ca, Si and Y, and at the ppm level for Al, Co, Cr, Fe, Hf, K, Li, Mn, Na, Ni, Se, Ti, V and Zr. Atomic emission spectroscopy with direct current are excitation and photographic detection has been employed. In order to eliminate the effect due to the differences in density between standards and samples, which are a source of errors, a chemical treatment of both is carried out. Likewise, for attaining conditions more suitable for the volatilization of certain impurities, these are determined with the sample in fluoride form. (Author) 11 refs

  9. Spectrographic determination of impurities in ceramic materials for nuclear fusion reactors III. Analysis of magnesium oxide

    International Nuclear Information System (INIS)

    Melon, A.M.; Roca, M.; Rucandio, M.I.

    1992-01-01

    The determination of minor and trace elements in the magnesium oxide, considered as possible ceramic material in thermonuclear fusion reactors, has been studied. The concentration ranges are 0.1 - 0.3 % for Ca, Si and Y, and at the ppm level for Al, Co, Cr, Fe, Hf, K, Li, Mn, Na, Ni, Sc, Ti, V and Zr. Atomic emission spectroscopy with direct current arc excitation and photographic detection has been employed. In order to eliminated the effect due to the differences in density between standards and samples, which are a source of errors, a chemical treatment of both is carried out. Likewise, for attaining conditions more suitable for the volatilization of certain impurities, these are determined with the sample in fluoride form. (author)

  10. Yield optimization in a cycled trickle-bed reactor: ethanol catalytic oxidation as a case study

    Energy Technology Data Exchange (ETDEWEB)

    Ayude, A.; Haure, P. [INTEMA, CONICET, Mar del Plata (Argentina); Cassanello, M. [Universidad de Buenos Aires, PINMATE, Departamento de Industrias, FCEyN, Buenos Aires (Argentina); Martinez, O. [Departamento de Ingenieria Quimica, FI-UNLP-CINDECA, La Plata (Argentina)

    2012-05-15

    The effect of slow ON-OFF liquid flow modulation on the yield of consecutive reactions is investigated for oxidation of aqueous ethanol solutions using a 0.5 % Pd/Al{sub 2}O{sub 3} commercial catalyst in a laboratory trickle-bed reactor. Experiments with modulated liquid flow rate (MLFR) were performed under the same hydrodynamic conditions (degree of wetting, liquid holdup) as experiments with constant liquid flow rate (CLFR). Thus, the impact of the duration of wet and dry cycles as well as the period can be independently investigated. Depending on cycling conditions, acetaldehyde or acetic acid production is favored with MLFR compared to CLFR. Results suggest both the opportunity and challenge of finding a way to tune the cycling parameters for producing the most appropriate product. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  11. Oscillations in the permanganate oxidation of glycine in a stirred flow reactor.

    Science.gov (United States)

    Poros, Eszter; Kurin-Csörgei, Krisztina; Szalai, István; Orbán, Miklós

    2013-09-19

    Oscillatory behavior is reported in the permanganate oxidation of glycine in the presence of Na2HPO4 in a stirred flow reactor. In near-neutral solutions, long-period sustained oscillations were recorded in the potential of a Pt electrode and in the light absorbance measured at λ = 418 and 545 nm, characteristic wavelengths for following the evolution of the intermediate [Mn(IV)] and reagent [MnO4(-) ] during the course of the reaction. No evidence of bistability was found. The chemical and physical backgrounds of the oscillatory phenomenon are discussed. In the oscillatory cycle, the positive feedback is attributed to the autocatalytic formation of a soluble Mn(IV) species, whereas the negative feedback arises from its removal from the solution in the form of solid MnO2. A simple model is suggested that qualitatively simulates the experimental observations in batch runs and the dynamics that appears in the flow system.

  12. Plasmachemical oxidation processes in a hybrid gas-liquid electrical discharge reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lukes, Petr; Locke, Bruce R [Department of Chemical and Biomedical Engineering, FAMU-FSU College of Engineering, Florida State University, 2525 Pottsdamer Street, Tallahassee, Florida (United States)

    2005-11-21

    Oxidation processes induced in water by pulsed electrical discharges generated simultaneously in the gas phase in close proximity to the water surface and directly in the liquid were investigated in a hybrid series gas-liquid electrical discharge reactor. The mechanism of phenol degradation was studied through its dependence on the gas phase and liquid phase compositions using pure argon and oxygen atmospheres above the liquid and different initial pH values in the aqueous solution. Phenol degradation was significantly enhanced in the hybrid-series reactor compared with the phenol removal by the single-liquid phase discharge reactor. Under an argon atmosphere the mechanism of phenol degradation was mainly caused by the electrophilic attack of OH{center_dot} radicals produced by the liquid phase discharge directly in water and OH{center_dot} radicals produced by the gas phase discharge at the gas-liquid interface. Under an oxygen atmosphere the formation of gaseous ozone dominated over the formation of OH{center_dot} radicals, and the contribution of the gas phase discharge in this case was determined mainly by the dissolution of gaseous ozone into the water and its subsequent interaction with phenol. At high pH phenol was degraded, in addition to the direct attack by ozone, also through indirect reactions of OH{center_dot} radicals formed via a peroxone process by the decomposition of dissolved ozone by hydrogen peroxide produced by the liquid phase discharge. Such a mechanism was proved by the detection of cis,cis-muconic acid and pH-dependent degradation of phenol, which resulted in significantly higher removal of phenol from alkaline solution observed under oxygen atmosphere than in argon.

  13. WATER-GAS SHIFT KINETICS OVER IRON OXIDE CATALYSTS AT MEMBRANE REACTOR CONDITIONS; A

    International Nuclear Information System (INIS)

    Carl R.F. Lund

    2001-01-01

    This report covers the second year of a project investigating water-gas shift catalysts for use in membrane reactors. It has been established that a simple iron high temperature shift catalyst becomes ineffective in a membrane reactor because the reaction rate is severely inhibited by the build-up of the product CO(sub 2). During the past year, an improved microkinetic model for water-gas shift over iron oxide was developed. Its principal advantage over prior models is that it displays the correct asymptotic behavior at all temperatures and pressures as the composition approaches equilibrium. This model has been used to explore whether it might be possible to improve the performance of iron high temperature shift catalysts under conditions of high CO(sub 2) partial pressure. The model predicts that weakening the surface oxygen bond strength by less than 5% should lead to higher catalytic activity as well as resistance to rate inhibition at higher CO(sub 2) partial pressures. Two promoted iron high temperature shift catalysts were studied. Ceria and copper were each studied as promoters since there were indications in the literature that they might weaken the surface oxygen bond strength. Ceria was found to be ineffective as a promoter, but preliminary results with copper promoted FeCr high temperature shift catalyst show it to be much more resistant to rate inhibition by high levels of CO(sub 2). Finally, the performance of sulfided CoMo/Al(sub 2)O(sub 3) catalysts under conditions of high CO(sub 2) partial pressure was simulated using an available microkinetic model for water-gas shift over this catalyst. The model suggests that this catalyst might be quite effective in a medium temperature water-gas shift membrane reactor, provided that the membrane was resistant to the H(sub 2)S that is required in the feed

  14. Plasmachemical oxidation processes in a hybrid gas-liquid electrical discharge reactor

    International Nuclear Information System (INIS)

    Lukes, Petr; Locke, Bruce R

    2005-01-01

    Oxidation processes induced in water by pulsed electrical discharges generated simultaneously in the gas phase in close proximity to the water surface and directly in the liquid were investigated in a hybrid series gas-liquid electrical discharge reactor. The mechanism of phenol degradation was studied through its dependence on the gas phase and liquid phase compositions using pure argon and oxygen atmospheres above the liquid and different initial pH values in the aqueous solution. Phenol degradation was significantly enhanced in the hybrid-series reactor compared with the phenol removal by the single-liquid phase discharge reactor. Under an argon atmosphere the mechanism of phenol degradation was mainly caused by the electrophilic attack of OH· radicals produced by the liquid phase discharge directly in water and OH· radicals produced by the gas phase discharge at the gas-liquid interface. Under an oxygen atmosphere the formation of gaseous ozone dominated over the formation of OH· radicals, and the contribution of the gas phase discharge in this case was determined mainly by the dissolution of gaseous ozone into the water and its subsequent interaction with phenol. At high pH phenol was degraded, in addition to the direct attack by ozone, also through indirect reactions of OH· radicals formed via a peroxone process by the decomposition of dissolved ozone by hydrogen peroxide produced by the liquid phase discharge. Such a mechanism was proved by the detection of cis,cis-muconic acid and pH-dependent degradation of phenol, which resulted in significantly higher removal of phenol from alkaline solution observed under oxygen atmosphere than in argon

  15. Control program of the neutron four-circle-diffractometer P32 at the SILOE reactor/CEN Grenoble

    International Nuclear Information System (INIS)

    Guth, H.; Paulus, H.; Reimers, W.; Heger, G.

    1983-09-01

    The four-circle diffractometer P32 for elastic neutron scattering on single crystals was installed at the SILOE reactor/CEN Grenoble in 1981. The control program, presented here, is a new update of the former program versions used at the FR2 reactor/Kernforschungszentrum Karlsruhe. Important improvements concerning reliability and handling of the diffractometer are added. (orig.) [de

  16. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  17. Modeling of simultaneous anaerobic methane and ammonium oxidation in a membrane biofilm reactor.

    Science.gov (United States)

    Chen, Xueming; Guo, Jianhua; Shi, Ying; Hu, Shihu; Yuan, Zhiguo; Ni, Bing-Jie

    2014-08-19

    Nitrogen removal by using the synergy of denitrifying anaerobic methane oxidation (DAMO) and anaerobic ammonium oxidation (Anammox) microorganisms in a membrane biofilm reactor (MBfR) has previously been demonstrated experimentally. In this work, a mathematical model is developed to describe the simultaneous anaerobic methane and ammonium oxidation by DAMO and Anammox microorganisms in an MBfR for the first time. In this model, DAMO archaea convert nitrate, both externally fed and/or produced by Anammox, to nitrite, with methane as the electron donor. Anammox and DAMO bacteria jointly remove the nitrite fed/produced, with ammonium and methane as the electron donor, respectively. The model is successfully calibrated and validated using the long-term (over 400 days) dynamic experimental data from the MBfR, as well as two independent batch tests at different operational stages of the MBfR. The model satisfactorily describes the methane oxidation and nitrogen conversion data from the system. Modeling results show the concentration gradients of methane and nitrogen would cause stratification of the biofilm, where Anammox bacteria mainly grow in the biofilm layer close to the bulk liquid and DAMO organisms attach close to the membrane surface. The low surface methane loadings result in a low fraction of DAMO microorganisms, but the high surface methane loadings would lead to overgrowth of DAMO bacteria, which would compete with Anammox for nitrite and decrease the fraction of Anammox bacteria. The results suggest an optimal methane supply under the given condition should be applied not only to benefit the nitrogen removal but also to avoid potential methane emissions.

  18. Participation in the United States Department of Energy Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.

  19. Participation in the United States Department of Energy Reactor Sharing Program

    International Nuclear Information System (INIS)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993

  20. A Mixed-Oxide Assembly Design for Rapid Disposition of Weapons Plutonium in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Adams, Marvin L.

    2002-01-01

    We have created a new mixed-oxide (MOX) fuel assembly design for standard pressurized water reactors (PWRs). Design goals were to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which we call MIX-33, offers some advantages for the disposition of weapons-grade plutonium; it increases the disposition rate by 8% while increasing the worth of control material, compared to a previous Westinghouse design. The MIX-33 design is based upon two ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the addition of water holes in the assembly. The main result of this paper is that both of these ideas are effective at increasing Pu throughput and increasing the worth of control material. With this new design, according to our analyses, we can transition smoothly from a full low-enriched-uranium (LEU) core to a full MIX-33 core while meeting the operational and safety requirements of a standard PWR. Given an interruption of the MOX supply, we can transition smoothly back to full LEU while meeting safety margins and using standard LEU assemblies with uniform pinwise enrichment distribution. If the MOX supply is interrupted for only one cycle, the transition back to a full MIX-33 core is not as smooth; high peaking could cause power to be derated by a few percent for a few weeks at the beginning of one transition cycle