WorldWideScience

Sample records for advanced gas-cooled nuclear

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-07

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-25

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-25

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-11-24

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

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

    Science.gov (United States)

    Mella, R.; Wenman, M. R.

    2013-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Marvin, M.D.

    1978-10-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  10. Advances in conceptual design of a gas-cooled accelerator driven system (ADS) transmutation devices to sustainable nuclear energy development

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, Rosales; Fajardo, Garcia; Curbelo, Perez; Oliva, Munoz; Hernandez, Garcia, E-mail: jrosales@instec.cu [Higher Institute of Technologies and Applied Sciences, Habana City (Cuba); Castells, Escriva [Energetic Engeniering Institute, Politechnical University of Valencia, Valencia (Spain); Abanades [Department of Simulation of Termoenergetic Systems, Politechnical University of Madrid, Madrid (Spain)

    2011-07-01

    The possibilities of a nuclear energy development are considerably increasing with the world energetic demand increment. However, the management of nuclear waste from conventional nuclear power plants and its inventory minimization are the most important issues that should be addressed. Fast reactors and Accelerator Driven Systems (ADS) are the main options to reduce the long-lived radioactive waste inventory. Pebble Bed Very High Temperature advanced systems have great perspectives to assume the future nuclear energy development challenges. The conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made in preliminary studies. The TADSEA is an ADS cooled by helium and moderated by graphite that uses as fuel small amounts of transuranic elements in the form of TRISO particles, confined in 3 cm radius graphite pebbles forming a pebble bed configuration. It would be used for nuclear waste transmutation and energy production. In this paper, the results of a method for calculating the number of whole pebbles fitting in a volume according to its size are showed. From these results, the packing fraction influence on the TADSEAs main work parameters is studied. In addition, a redesign of the previous configuration, according to the established conditions in the preliminary design, i.e. the exit thermal power, is made. On the other hand, the heterogeneity of the TRISO particles inside the pebbles can not be negligible. In this paper, a study of the power density distribution inside the pebbles by means of a detailed simulation of the TRISO fuel particles and using an homogeneous composition of the fuel is addressed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-11-14

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-12

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

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

    Science.gov (United States)

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

    2011-03-01

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

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

    Science.gov (United States)

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

    2013-09-03

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

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

    Science.gov (United States)

    Palmer, David J.

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

  17. Application of gas-cooled Accelerator Driven System (ADS) transmutation devices to sustainable nuclear energy development

    OpenAIRE

    2011-01-01

    The conceptual design of a pebble bed gas-cooled transmutation device is shown with the aim to evaluate its potential for its deployment in the context of the sustainable nuclear energy development, which considers high temperature reactors for their operation in cogeneration mode, producing electricity, heat and Hydrogen. As differential characteristics our device operates in subcritical mode, driven by a neutron source activated by an accelerator that adds clear safety advantages and fuel f...

  18. A small high temperature gas cooled reactor for nuclear marine propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Brugiere, F.; Sillon, C. [Ecole des Applications Militaires de l' Energie Atomique, 50 - Cherbourg (France); Foster, A.; Hamilton, P.; Jewer, S.; Thompson, A.C. [Defence College of Electromechanical Engineering, Nuclear Dept., Military Rd, Gosport (United Kingdom); Kingston, T.; Williams, A.M.; Beeley, P.A. [Rolls-Royce (Marine Power), Raynesway, Derby (United Kingdom)

    2007-07-01

    Results from a design study for a hypothetical nuclear marine propulsion plant are presented. The plant utilizes a small High Temperature Gas Cooled Reactor (HTGCR) similar to the GTHTR300 design by the Japan Atomic Energy Agency with power being generated by a direct cycle gas turbine. The GTHTR300 design is modified in order to achieve the required power of 80 MWth and core lifetime of approximately 10 years. Thermal hydraulic analysis shows that in the event of a complete loss of flow accident the hot channel fuel temperature exceeds the 1600 Celsius degrees limit due to the high power peaking in assemblies adjacent to the inner reflector. Reactor dynamics shows oscillatory behaviour in rapid power transients. An automatic control rod system is suggested to overcome this problem. (authors)

  19. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  20. Analysis of the horizontal flow in the advanced gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Y. [Department of Mechanical Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); He, S., E-mail: s.he@sheffield.ac.uk [Department of Mechanical Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Ganesan, P. [Department of Mechanical Engineering, University of Malaya, Kuala Lumpur 50603 (Malaysia); Gotts, J. [EDF Energy, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2014-06-01

    Highlights: • CFD is used to assess the effect of horizontal flows in AGRs. • The horizontal flows can reduce the graphite brick temperature significantly. • Such effects are not taken into consideration in current engineering calculations. • There might be flow instabilities when the fuel channel flow is very low but horizontal flows reduce its possibility. - Abstract: The purpose of the paper is to report a computational investigation of horizontal flows in the UK advanced-gas-cooled reactor (AGR) by using computational fluid dynamics with ANSYS FLUENT. The study is relevant to practical issues encountered in some AGR stations currently in operation in the UK. It is carried out using a comparative approach based on the results of two contrasting models: one simulating the full effect of the cross flow, the other simulating the simplified approach currently employed by the industry which neglects the momentum of the horizontal cross flow. The study reveals that the horizontal cross flow plays a significant role in the cooling of the moderator brick, while the axial variation of the brick geometry also significantly changes the distribution of the temperature within the brick. It is also found that under some circumstances the so-called horizontal inter-brick leakage (HIBL) flow could influence the cooling performance in the narrow gaps, resulting in a local hot spot. Furthermore, there may be flow instabilities in the flows in AGR fuel channels due to the interactions between the flow in the main arrowhead flow passages and that in some narrow passages connected to it, but the influence on the brick temperature is negligible. Horizontal cross flow has an effect of reducing such instabilities.

  1. Safety assessment for electricity generation failure accident of gas cooled nuclear power plant using system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    2013-04-15

    The power production failure happens in the loss of coolant of the nuclear power plants (NPPs). The air ingress is a serious accident in gas cooled NPPs. The quantification of the study performed by the system dynamics (SD) method which is processed by the feedback algorithms. The Vensim software package is used for the simulation, which is performed by the Monte-Carlo method. Two kinds of considerations as the economic and safety properties are important in NPPs. The result shows the stability of the operation when the power can be decided. The maximum value of risk is the 11.77 in 43rd and the minimum value is 0.0 in several years. So, the success of the circulation of coolant is simulated by the dynamical values. (orig.)

  2. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  3. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  4. Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)

    2011-07-01

    The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which

  5. A History of Dosimetry for the Advanced Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    Shaw Simon

    2016-01-01

    Full Text Available This paper presents a summary of the methods used in the first ∼40 years of AGR neutron dosimetry and nuclear heating calculations, and the influence of the earlier Magnox reactor dosimetry programme. While the current state-of-the-art Monte Carlo methods are extremely powerful they still require very careful consideration of the quality of the input data, nuclear data validation and variance reduction techniques; in particular, this paper examines the difficulties in assuring the adequate convergence of calculations when Monte Carlo acceleration is applied in the presence of significant streaming paths through attenuating or scattering media.

  6. A History of Dosimetry for the Advanced Gas-cooled Reactors

    Science.gov (United States)

    Shaw, Simon; Thornton, Dean

    2016-02-01

    This paper presents a summary of the methods used in the first ˜40 years of AGR neutron dosimetry and nuclear heating calculations, and the influence of the earlier Magnox reactor dosimetry programme. While the current state-of-the-art Monte Carlo methods are extremely powerful they still require very careful consideration of the quality of the input data, nuclear data validation and variance reduction techniques; in particular, this paper examines the difficulties in assuring the adequate convergence of calculations when Monte Carlo acceleration is applied in the presence of significant streaming paths through attenuating or scattering media.

  7. CFD-DEM simulation of a conceptual gas-cooled fluidized bed nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Lucilla C.; Su, Jian, E-mail: lucillalmeida@gmail.com, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Aguirre, Joao, E-mail: aguirre@rocky-dem.com [Engineering Simulation and Scientific Software (ESSS), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Several conceptual designs of the fluidized-bed nuclear reactor have been proposed due to its many advantages over conventional nuclear reactors such as PWRs and BWRs. Amongst their characteristics, the enhanced heat transfer and mixing enables a more uniform temperature distribution, reducing the risk of hot-spot and excessive fuel temperature, in addition to resulting in a higher burnup of the fuel. Furthermore, the relationship between the bed height and reactor neutronics turns the coolant flow rate control into a power production mechanism. Moreover, the possibility of removing the fuel by gravity from the movable core in case of a loss-of-cooling accident increases its safety. High-accuracy modeling of particles and coolant flow in fluidized bed reactors is needed to evaluate reliably the thermal-hydraulic efficiency and safety margin. The two-way coupling between solid and fluid can account for high-fidelity solid-solid interaction and reasonable accuracy in fluid calculation and fluid-solid interaction. In the CFD-DEM model, the particles are modeled as a discrete phase, following the DEM approach, whereas the fluid flow is treated as a continuous phase, described by the averaged Navier-Stokes equations on a computational cell scale. In this work, the coupling methodology between Fluent and Rocky is described. The numerical approach was applied to the simulation of a bubbling fluidized bed and the results were compared to experimental data and showed good agreement. (author)

  8. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  9. A simplified Probabilistic Safety Assesment of a Steam-Methane Reforming Hydrogen Production Plant coupled to a High-Temperature Gas Cooled Nuclear Reactor

    OpenAIRE

    Nelson Edelstein, Pamela; Flores Flores, Alain; Francois Lacouture, Juan Luis

    2005-01-01

    A Probabilistic Safety Assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a High-Temperature Gas Cooled Nuclear Reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute’s (JAERI) High Temperature Test Reactor (HTTR) prototype in Japan. This study has two major objectives: calculate the risk to onsite and offsite individuals, and calculate the frequency of different types of damage to the complex. A simplified HAZOP...

  10. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  11. Generation IV nuclear energy system initiative. Large GFR core subassemblydesign for the Gas-Cooled Fast Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, E. A.; Kulak, R. F.; Therios, I. U.; Wei, T. Y. C.

    2006-07-31

    Gas-cooled fast reactor (GFR) designs are being developed to meet Gen IV goals of sustainability, economics, safety and reliability, and proliferation resistance and physical protection as part of an International Generation IV Nuclear Energy System Research Initiative effort. Different organizations are involved in the development of a variety of GFR design concepts. The current analysis has focused on the evaluation of low-pressure drop, pin-core designs with favorable passive cooling properties. Initial evaluation of the passive cooling safety case for the GFR during depressurized decay heat removal accidents with concurrent loss of electric power have resulted in requirements for a reduction of core power density to the 100 w/cc level and a low core pressure drop of 0.5 bars. Additional design constraints and the implementation of their constraints are evaluated in this study to enhance and passive cooling properties of the reactor. Passive cooling is made easier by a flat radial distribution of the decay heat. One goal of this study was to evaluate the radial power distribution and determine to what extent it can be flattened, since the decay heat is nearly proportional to the fission power at shutdown. In line with this investigation of the radial power profile, an assessment was also made of the control rod configuration. The layout provided a large number of control rod locations with a fixed area provided for control rods. The number of control rods was consistent with other fast reactor designs. The adequacy of the available control rod locations was evaluated. Future studies will be needed to optimize the control rod designs and evaluate the shutdown system. The case for low pressure drop core can be improved by the minimization of pressure drop sources such as the number of required fuel spacers in the subassembly design and by the details of the fuel pin design. The fuel pin design is determined by a number of neutronic, thermal-hydraulic (gas dynamics

  12. Advances in nuclear science and technology

    CERN Document Server

    Greebler, Paul

    1968-01-01

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

  13. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  14. Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO{sub 2} emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO{sub 2} emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus{sup TM} Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition.

  15. A Critical Review of the Recent Improvements in Minimizing Nuclear Waste by Innovative Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    E. Bomboni

    2008-01-01

    Full Text Available This paper presents a critical review of the recent improvements in minimizing nuclear waste in terms of quantities, long-term activities, and radiotoxicities by innovative GCRs, with particular emphasis to the results obtained at the University of Pisa. Regarding these last items, in the frame of some EU projects (GCFR, PUMA, and RAPHAEL, we analyzed symbiotic fuel cycles coupling current LWRs with HTRs, finally closing the cycle by GCFRs. Particularly, we analyzed fertile-free and Pu-Th-based fuel in HTR: we improved plutonium exploitation also by optimizing Pu/Th ratios in the fuel loaded in an HTR. Then, we chose GCFRs to burn residual MA. We have started the calculations on simplified models, but we ended them using more “realistic” models of the reactors. In addition, we have added the GCFR multiple recycling option using keff calculations for all the reactors. As a conclusion, we can state that, coupling HTR with GCFR, the geological disposal issues concerning high-level radiotoxicity of MA can be considerably reduced.

  16. Dimethylamine as a replacement for ammonia dosing in the secondary circuit of an advanced gas-cooled reactor (AGR) power station

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, Chris; Mitchell, Malcolm S. [EDF Energy, Hartlepool Power Station, Hartlepool (United Kingdom); Bull, Andrew E.A.; Quirk, Graham P.; Rudge, Andy [EDF Energy Nuclear Generation, Barnwood, Gloucester (United Kingdom). Central Technical Organisation; Woolsey, Ian S.

    2012-06-15

    Increasing flow resistance observed over recent years within the helical once-through boilers in the four advanced gas-cooled reactors at Hartlepool and Heysham 1 Power Stations have reduced boiler performance, resulting in reductions in feedwater flow, steam temperatures, and power output and in the need to carry out periodic chemical cleaning. The root cause is believed to be the development of magnetite deposits with high flow impedance in the 9Cr1Mo evaporator section of the boiler tubing. To prevent continued increases in boiler flow resistance, dimethylamine is being trialled, in one of the four affected units, as a replacement to the conventional ammonia dosing. Dimethylamine increases the pH at temperature around the secondary circuit and, based on full scale boiler rig simulations, is expected to reduce iron transport and prevent flow resistance increases within the evaporator section of the boiler. The dimethylamine plant trial commenced in January 2011 and is ongoing. The feedwater concentration of dimethylamine has been increased progressively towards a final target value of 900 {mu}g . kg{sup -1} and its effect on iron transport and boiler pressure loss is being closely monitored. The high steam temperature (> 500 C) of the secondary circuit leads to some decomposition of dimethylamine, which is being carefully monitored at various locations around the circuit. The decomposition products identified with dimethylamine dosing include ammonia, methylamine, formic acid, carbon dioxide and, as yet, unidentified neutral organic species. The effect of dimethylamine dosing on iron transport and boiler pressure drops and its decomposition behaviour around the secondary circuit during the plant trial will be presented in this paper. (orig.)

  17. Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Roguin, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de

  18. 四代核电技术(高温堆)产业化探讨%Chinese Development of Generation IV Nuclear Power Technology--High Temperature Gas Cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    石磊; 肖国平; 鲁盛会; 张见营; 王芝芬; 李金英

    2015-01-01

    该文综述了四代核电技术(高温堆)的发展现状和技术特点,对中国高温气冷堆的产业化提出了建议,供政府及相关行业集团公司等参考。%The author reviewed the development of generation IV nuclear power technology --Temperature Gas Cooled Reactor(HTR) . Some The advice of industrialization of HTR in China was proposed.The article can give some advice for the government and relevant company.

  19. 200MW高温气冷堆汽轮机热力系统能损分析%Energy Loss Analysis of Turbine in 200MW High Temperature Gas Cooled Reactor Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    杨宇

    2015-01-01

    采用能级效率法对200 MW高温气冷堆核电机组的热力系统能损分析,特别是对各级加热器的能损进行了解耦分析.通过引入加热器的热耗影响因数,获得了VWO、TRL、75%TRL、50%TRL 4种工况下各级加热器对降低热力系统热耗的影响和变化规律.以上分析方法和结果,可以为200MW高温气冷堆核电机组的热力系统的设计、优化、运行和维护提供重要参考.%In this paper,energy loss analysis of thermal power system of 200MW high temperature gas cooled reactor nuclear power plant with the energy level efficiency method, especially for the energy loss decoupling analysis of each heater.By introducing the heat consumption influence coefficient, the influence of each heater in reducing the heat consumption of thermodynamic system under the conditions of VWO,TRL,75% ofTRL and 50% ofTRL was obtained.The above results provide important reference for the design,optimization,operation and maintenance of the thermal system of 200MW high temperature gas cooled reactor nuclear power plant.

  20. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  1. Advances in Nuclear Energy

    Science.gov (United States)

    Frois, B.

    2005-04-01

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

  2. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1982-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-18

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-06-26

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

  6. Alpha particle spectroscopy — A useful tool for the investigation of spent nuclear fuel from high temperature gas-cooled reactors

    Science.gov (United States)

    Helmbold, M.

    1984-06-01

    For more than a decade, alpha particle spectrometry of spent nuclear fuel has been used at the Kernforschungsanlage Jülich (KFA) in the field of research for the German high temperature reactor (HTR). Techniques used for the preparation of samples for alpha spectrometry have included deposition from aqueous solutions of spent fuel, annealing of fuel particles in an oven and the evaporation of fuel material by a laser beam. The resulting sources are very thin but of low activity and the alpha spectrometry data obtained from them must be evaluated with sophisticated computer codes to achieve the required accuracy. Measurements have been made on high and low enriched uranium fuel and on a variety of parameters relevant to the fuel cycle. In this paper the source preparation and data evaluation techniques will be discussed together with the results obtained to data, i.e. production of alpha active actinide isotopes, correlations between actinide isotopes and fission products, build up and transmutation of actinides during burn-up of HTR fuel, diffusion coefficients of actinides for fuel particle kernels and coating materials. All these KFA results have helped to establish the basis for the design, licensing and operation of HTR power plants, including reprocessing and waste management.

  7. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: reprocessing of high-temperature gas-cooled reactor fuel containing U-233 and thorium

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. Jr.; Blanco, R.E.; Finney, B.C.; Hill, G.S.; Moore, R.E.; Witherspoon, J.P.

    1976-05-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model high-temperature gas-cooled reactor (HTGR) fuel reprocessing plant and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist the U. S. Nuclear Regulatory Commission in defining the term as low as reasonably achievable as it applies to this nuclear facility. The base case is representative of conceptual, developing technology of head-end graphite-burning operations and of extensions of solvent-extraction technology of current designs for light-water-reactor (LWR) fuel reprocessing plants. The model plant has an annual capacity of 450 metric tons of heavy metal (MTHM, where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods used in the case studies is discussed.

  8. Computational fluid dynamics analysis of core bypass flow and crossflow in a prismatic very high temperature gas-cooled nuclear reactor based on a two-layer block model

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Huhu, E-mail: huhuwang@tamu.edu [Department of Nuclear Engineering, Texas A and M University, 3133 TAMU, College Station, TX 77840 (United States); Dominguez-Ontiveros, Elvis, E-mail: elvisdom@tamu.edu [Department of Nuclear Engineering, Texas A and M University, 3133 TAMU, College Station, TX 77840 (United States); Hassan, Yassin A., E-mail: y-hassan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, 3133 TAMU, College Station, TX 77840 (United States); Department of Mechanical Engineering, Texas A and M University, 3123 TAMU, College Station, TX 77840 (United States)

    2014-03-15

    Highlights: • A CFD model was built based on a two-layer block experimental facility at Texas A and M University. • The coolant characterizations within the uniform and wedge-shaped crossflow gap regions were investigated. • The influence on the coolant distribution from the bypass flow gap width was studied. • Discretization and iterative errors involved in the simulations were quantified. - Abstract: The very high temperature gas-cooled nuclear reactor (VHTR) has been designated as one of the promising reactors that will serve for the Next Generation (Generation IV) Nuclear Plant. For a prismatic VHTR core, the bypass flow and crossflow phenomena are important design considerations. To investigate the coolant distribution in the reactor core based on the two-layer block facility built at Texas A and M University, a three-dimensional steady-state CFD analysis was performed using the commercial code STAR-CCM+ v6.04. Results from this work serve as a guideline and validating source for the related experiments. A grid independence study was conducted to quantify related errors in the simulations. The simulation results show that the bypass flow fraction was not a strong function of the Reynolds number. The presence of the crossflow gap had a significant effect on the distribution of the coolant in the core. Uniform and wedge-shape crossflow gaps were studied. It was found that a significant secondary flow in the crossflow gap region moved from the bypass flow gap toward coolant holes, which resulted in up to a 28% reduction of the coolant mass flow rate in the bypass flow gap.

  9. Fuel development for gas-cooled fast reactors

    Science.gov (United States)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  10. Production of Depleted UO2Kernels for the Advanced Gas-Cooled Reactor Program for Use in TRISO Coating Development

    Energy Technology Data Exchange (ETDEWEB)

    Collins, J.L.

    2004-12-02

    The main objective of the Depleted UO{sub 2} Kernels Production Task at Oak Ridge National Laboratory (ORNL) was to conduct two small-scale production campaigns to produce 2 kg of UO{sub 2} kernels with diameters of 500 {+-} 20 {micro}m and 3.5 kg of UO{sub 2} kernels with diameters of 350 {+-} 10 {micro}m for the U.S. Department of Energy Advanced Fuel Cycle Initiative Program. The final acceptance requirements for the UO{sub 2} kernels are provided in the first section of this report. The kernels were prepared for use by the ORNL Metals and Ceramics Division in a development study to perfect the triisotropic (TRISO) coating process. It was important that the kernels be strong and near theoretical density, with excellent sphericity, minimal surface roughness, and no cracking. This report gives a detailed description of the production efforts and results as well as an in-depth description of the internal gelation process and its chemistry. It describes the laboratory-scale gel-forming apparatus, optimum broth formulation and operating conditions, preparation of the acid-deficient uranyl nitrate stock solution, the system used to provide uniform broth droplet formation and control, and the process of calcining and sintering UO{sub 3} {center_dot} 2H{sub 2}O microspheres to form dense UO{sub 2} kernels. The report also describes improvements and best past practices for uranium kernel formation via the internal gelation process, which utilizes hexamethylenetetramine and urea. Improvements were made in broth formulation and broth droplet formation and control that made it possible in many of the runs in the campaign to produce the desired 350 {+-} 10-{micro}m-diameter kernels, and to obtain very high yields.

  11. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: fabrication of high-temperature gas-cooled reactor fuel containing uranium-233 and thorium

    Energy Technology Data Exchange (ETDEWEB)

    Roddy, J.W.; Blanco, R.E.; Hill, G.S.; Moore, R.E.; Seagren, R.D.; Witherspoon, J.P.

    1976-06-01

    A cost/benefit study was made to determine the cost and effectiveness of various radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from model High-Temperature Gas-Cooled (HTGR) fuel fabrication plants and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as reasonably achievable'' as it applies to these nuclear facilities. The base cases of the two model plants, a fresh fuel fabrication plant and a refabrication plant, are representative of current proposed commercial designs or are based on technology that is being developed to fabricate uranium, thorium, and graphite into fuel elements. The annual capacities of the fresh fuel plant and the refabrication plant are 450 and 245 metric tons of heavy metal (where heavy metal is uranium plus thorium), as charged to about fifty 1000-MW(e) HTGRs. Additional radwaste treatment systems are added to the base case plants in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The capital and annual costs for the added waste treatment operations and the corresponding reductions in dose commitments are calculated for each case. In the final analysis, the cost/benefit of each case, calculated as additional cost of radwaste system divided by the reduction in dose commitment, is tabulated or the dose commitment is plotted with cost as the variable. The status of each of the radwaste treatment methods is discussed. 48 figures, 74 tables.

  12. Advances in nuclear physics

    CERN Document Server

    Vogt, Erich

    1975-01-01

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

  13. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  14. Recent advances in nuclear cardiology

    DEFF Research Database (Denmark)

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

    2008-01-01

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

  15. Advanced nuclear precleaner

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

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

  16. Recent advances in nuclear cardiology

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  17. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  18. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  19. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

  20. Gas-cooled reactors: the importance of their development

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U/sub 3/O/sub 8/ before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production.

  1. A buffer gas cooled beam of barium monohydride

    Science.gov (United States)

    Iwata, Geoffrey; Tarallo, Marco; Zelevinsky, Tanya

    2016-05-01

    Significant advances in direct laser cooling of diatomic molecules have opened up a wide array of molecular species to precision studies spanning many-body physics, quantum collisions and ultracold dissociation. We present a cryogenic beam source of barium monohydride (BaH), and study laser ablation of solid precursor targets as well as helium buffer gas cooling dynamics. Additionally, we cover progress towards a molecular magneto-optical trap, with spectroscopic studies of relevant cooling transitions in the B2 Σ molecules, including resolution of hyperfine structure and precision measurements of the vibrational Frank-Condon factors. Finally, we examine the feasibility of photo dissociation of trapped BaH molecules to yield optically accessible samples of ultracold hydrogen.

  2. A gas-cooled reactor surface power system

    Science.gov (United States)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  3. Application of Hastelloy X in Gas-Cooled Reactor Systems

    DEFF Research Database (Denmark)

    Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W.R.

    1976-01-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data...

  4. Advanced nuclear thermal propulsion concepts

    Science.gov (United States)

    Howe, Steven D.

    1993-11-01

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

  5. Advanced nuclear thermal propulsion concepts

    Science.gov (United States)

    Howe, Steven D.

    1993-01-01

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

  6. Efficiency and cost advantages of an advanced-technology nuclear electrolytic hydrogen-energy production facility

    Science.gov (United States)

    Donakowski, T. D.; Escher, W. J. D.; Gregory, D. P.

    1977-01-01

    The concept of an advanced-technology (viz., 1985 technology) nuclear-electrolytic water electrolysis facility was assessed for hydrogen production cost and efficiency expectations. The facility integrates (1) a high-temperature gas-cooled nuclear reactor (HTGR) operating a binary work cycle, (2) direct-current (d-c) electricity generation via acyclic generators, and (3) high-current-density, high-pressure electrolyzers using a solid polymer electrolyte (SPE). All subsystems are close-coupled and optimally interfaced for hydrogen production alone (i.e., without separate production of electrical power). Pipeline-pressure hydrogen and oxygen are produced at 6900 kPa (1000 psi). We found that this advanced facility would produce hydrogen at costs that were approximately half those associated with contemporary-technology nuclear electrolysis: $5.36 versus $10.86/million Btu, respectively. The nuclear-heat-to-hydrogen-energy conversion efficiency for the advanced system was estimated as 43%, versus 25% for the contemporary system.

  7. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  8. A modular gas-cooled cermet reactor system for planetary base power

    Science.gov (United States)

    Jahshan, Salim N.; Borkowski, Jeffrey A.

    1993-01-01

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. A six module gas-cooled, cermet-fueled reactor is proposed that can meet the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  9. Helium mobility in advanced nuclear ceramics

    OpenAIRE

    Agarwal, Shradha

    2014-01-01

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

  10. Evaluating the income and employment impacts of gas cooling technologies

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, P.J. [Oak Ridge National Lab., TN (United States); Laitner, S.

    1995-03-01

    The purpose of this study is to estimate the potential employment and income benefits of the emerging market for gas cooling products. The emphasis here is on exports because that is the major opportunity for the U.S. heating, ventilating, and air-conditioning (HVAC) industry. But domestic markets are also important and considered here because without a significant domestic market, it is unlikely that the plant investments, jobs, and income associated with gas cooling exports would be retained within the United States. The prospects for significant gas cooling exports appear promising for a variety of reasons. There is an expanding need for cooling in the developing world, natural gas is widely available, electric infrastructures are over-stressed in many areas, and the cost of building new gas infrastructure is modest compared to the cost of new electric infrastructure. Global gas cooling competition is currently limited, with Japanese and U.S. companies, and their foreign business partners, the only product sources. U.S. manufacturers of HVAC products are well positioned to compete globally, and are already one of the faster growing goods-exporting sectors of the U.S. economy. Net HVAC exports grew by over 800 percent from 1987 to 1992 and currently exceed $2.6 billion annually (ARI 1994). Net gas cooling job and income creation are estimated using an economic input-output model to compare a reference case to a gas cooling scenario. The reference case reflects current policies, practices, and trends with respect to conventional electric cooling technologies. The gas cooling scenario examines the impact of accelerated use of natural gas cooling technologies here and abroad.

  11. Description of the magnox type of gas cooled reactor (MAGNOX)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO{sub 2}) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kimball, O.F.

    1982-06-15

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

  13. Advanced-gas-cooled-nuclear-reactor materials evaluation and development program. Volume 1. Final report, September 23, 1976-September 30, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kimball, O.F.

    1983-05-15

    Included in this report is a discussion of the materials selected for the screening phase and more intensive screening phase test programs and the systems and components for which they are candidate materials. Thirty-one (31) commercially available alloy and alloy/coating materials and ten (10) experimental alloys were evaluated in the program. The experimental test facilities developed as part of this program are discussed and experimental testing procedures are summarized. The results of the initial screening test programs are presented. This includes creep testing results and metallographic analyses of candidate materials exposed to simulated HTGR helium and air under stress at temperatures of 750/sup 0/, 850/sup 0/, 950/sup 0/, or 1050/sup 0/C (1382/sup 0/, 1562/sup 0/, 1742/sup 0/, or 1922/sup 0/F) for exposure times to 10,000 hours. Metallographic analyses, weight change and carbon analyses results, and post exposure room temperature tensile and Charpy V-notch impact test results are presented for candidate materials exposed unstressed under the conditions stated above.

  14. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  15. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1962-01-01

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

  16. Nuclear Thermal Propulsion for Advanced Space Exploration

    Science.gov (United States)

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

    2012-01-01

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

  17. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Lap-Yan Cheng

    2009-01-01

    Full Text Available The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR in a GEN IV direct-cycle gas-cooled fast reactor (GFR which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  18. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  19. Advances in nuclear science and technology

    CERN Document Server

    Greebler, Paul

    1966-01-01

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

  20. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1970-01-01

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

  1. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  2. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1972-01-01

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

  3. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1976-01-01

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

  4. A preliminary systems-engineering study of an advanced nuclear-electrolytic hydrogen-production facility

    Science.gov (United States)

    Escher, W. J. D.; Donakowski, T. D.; Tison, R. R.

    1975-01-01

    An advanced nuclear-electrolytic hydrogen-production facility concept was synthesized at a conceptual level with the objective of minimizing estimated hydrogen-production costs. The concept is a closely-integrated, fully-dedicated (only hydrogen energy is produced) system whose components and subsystems are predicted on ''1985 technology.'' The principal components are: (1) a high-temperature gas-cooled reactor (HTGR) operating a helium-Brayton/ammonia-Rankine binary cycle with a helium reactor-core exit temperature of 980 C, (2) acyclic d-c generators, (3) high-pressure, high-current-density electrolyzers based on solid-polymer electrolyte technology. Based on an assumed 3,000 MWt HTGR the facility is capable of producing 8.7 million std cu m/day of hydrogen at pipeline conditions, 6,900 kPa. Coproduct oxygen is also available at pipeline conditions at one-half this volume. It has further been shown that the incorporation of advanced technology provides an overall efficiency of about 43 percent, as compared with 25 percent for a contemporary nuclear-electric plant powering close-coupled contemporary industrial electrolyzers.

  5. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

    This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.

  6. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  8. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1973-01-01

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

  9. Advanced research workshop: nuclear materials safety

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J; Moshkov, M M

    1999-01-28

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

  10. Development status and operational features of the high temperature gas-cooled reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Winkleblack, R.K.

    1976-04-01

    The objective of this study is to investigate the maturity of HTR-technology and to look out for possible technical problems, concerning introduction of large HTR power plants into the market. Further state and problems of introducing and closing the thorium fuel cycle is presented and judged. Finally, the state of development of advanced HTR-concepts for electricity production, the direct cycle HTR with helium turbine, and the gas-cooled fast breeder is discussed. In preparing the study, both HTR concepts with spherical and block-type fuel elements have been considered.

  11. Advances in nuclear science and technology

    CERN Document Server

    Henley, Ernest J

    1975-01-01

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

  12. Computational Design of Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-03

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

  13. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  14. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  15. Recent Advances in Ocean Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Kang-Heon Lee

    2015-10-01

    Full Text Available In this paper, recent advances in Ocean Nuclear Power Plants (ONPPs are reviewed, including their general arrangement, design parameters, and safety features. The development of ONPP concepts have continued due to initiatives taking place in France, Russia, South Korea, and the United States. Russia’s first floating nuclear power stations utilizing the PWR technology (KLT-40S and the spar-type offshore floating nuclear power plant designed by a research group in United States are considered herein. The APR1400 and SMART mounted Gravity Based Structure (GBS-type ONPPs proposed by a research group in South Korea are also considered. In addition, a submerged-type ONPP designed by DCNS of France is taken into account. Last, issues and challenges related to ONPPs are discussed and summarized.

  16. Advanced nuclear reactor public opinion project

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

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

  17. Experimental and numerical investigation of heat transfer and pressure drop for innovative gas cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, R., E-mail: rodrigo.leija@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz No. 1, 76344 Eggenstein-Leopoldshafen (Germany); Buchholz, S. [Gesellschaft für Anlagen- und Reaktorsicherheit GRS mbH, Boltzmannstraße 2, 85748 Garching (Germany); Suikkanen, H. [Lappeenranta University of Technology, LUT Energy, PO Box 20, FI-53851 Lappeenranta (Finland)

    2015-08-15

    Highlights: • Experimental results of the L-STAR within the first stage of THINS project. • CFD validation for the heat transfer and pressure losses in innovative gas cooled systems. • The results indicate a strong dependency Turbulent Prandtl at the rod wall temperature distribution. • Gas loop facility suitable for the investigation of thermohydraulic issues of GFR, however there might be flow instabilities when flow is very low. - Abstract: Heat transfer enhancement through turbulence augmentation is recognized as a key factor for improving the safety and economic conditions in the development of both critical and subcritical innovative advanced gas cooled fast reactors (GFR) and transmutation systems. The L-STAR facility has been designed and erected at the Karlsruhe Institute of Technology (KIT) to study turbulent flow behavior and its heat transfer enhancement characteristics in gas cooled annular channels under a wide range of conditions. The test section consists of an annular hexagonal cross section channel with an inner electrical heater rod element, placed concentrically within the test section, which seeks to simulate the flow area of a fuel rod element in a GFR. The long term objective of the experimental study is to investigate and improve the understanding of complex turbulent convective enhancement mechanisms as well as the friction loss penalties of roughened fuel rods compared to smooth ones and to generate an accurate database for further development of physical models. In the first step, experimental results of the fluid flow with uniform heat release conditions for the smooth heater rod are presented. The pressure drops, as well as the axial temperature profiles along the heater rod surface have been measured at Reynolds numbers in the range from 4000 to 35,000. The experimental results of the first stage were compared with independently conducted CFD analyses performed at Lappeenranta University of Technology (LUT) with the code ANSYS

  18. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  19. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

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

    Directory of Open Access Journals (Sweden)

    Wang Qiu-kuan

    2012-01-01

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

  1. Advancing Underground Nuclear Astrophysics with CASPAR

    Science.gov (United States)

    Robertson, Daniel; Couder, Manoel; Greife, Uwe; Strieder, Frank; Wells, Doug; Wiescher, Michael

    2015-04-01

    The advancement of experimental nuclear astrophysics techniques and the requirement of astrophysical network models for further nuclear data over greater energy ranges, has led to the requirement for the better understanding of nuclear reactions in stellar burning regimes. For those reactions of importance to stellar burning processes and elemental production through stellar nucleosynthesis, the energy range of astrophysical interest is always problematic to probe. As reaction measurements approach the burning window of interest, the rapid drop off in cross-section hampers laboratory investigation. The natural background suppression of underground accelerator facilities enables the extension of current experimental data to lower energies. An example of such reactions of interest are those thought to be sources of neutrons for the s-process, the major production mechanism for elements above the iron peak. The reactions 13 C(α,n)16 O and 22 Ne(α,n)25 Mg are the proposed initial focus of the new nuclear astrophysics accelerator laboratory (CASPAR) currently under construction at the Sanford Underground Research Facility, Lead, SD. With thanks to funding provided by South Dakota Science and Technology Authority and the NSF under Grant Number PHY-1419765.

  2. Concept of an inherently-safe high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro [Nuclear Hydrogen and Heat Application Research Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

    2012-06-06

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  3. Measurement of Turbulent Flow Phenomena for the Lower Plenum of a Prismatic Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink; Keith G. Condie; Glenn E. McCreery

    2007-09-01

    Mean velocity field and turbulence data are presented for flow phenomena in a lower plenum of a typical prismatic gas-cooled reactor (GCR), such as in a Very High Temperature Reactor (VHTR) concept. In preparation for design, safety analyses and licensing, research has begun on readying the computational tools that will be needed to predict the thermal-hydraulics behavior of the reactor design. Fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of computational fluid dynamics (CFD) codes and their turbulence models for a typical VHTR plenum geometry in the limiting case of negligible buoyancy and constant fluid properties. This experiment has been proposed as a “Standard Problem” for assessing advanced reactor (CFD) analysis tools. Present results concentrate on the region of the plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum can locally be considered as multiple jets into a confined cross flow - with obstructions. A model of the lower plenum has been fabricated and scaled to the geometric dimensions of the Next Generation Nuclear Plant (NGNP) Point Design. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to induce flow features somewhat comparable to those expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive-index of the working fluid so that optical techniques may be employed for the measurements. The experiments were conducted in the Matched-Index-of-Refraction (MIR) Facility at the Idaho National Laboratory (INL). The benefit of the MIR technique is that it permits optical measurements to determine complex flow characteristics in passages and around objects to be obtained without locating a disturbing transducer in the flow field and without distortion of the optical paths. The

  4. Heat transfer problems in gas-cooled solid blankets

    Energy Technology Data Exchange (ETDEWEB)

    Fillo, J.A.; Powell, J.R.

    1976-01-01

    In all fusion reactors using the deuterium-tritium fuel cycle, a large fraction approximately 80 percent of the fusion energy will be released as approximately 14 MeV neutrons which must be slowed down in a relatively thick blanket surrounding the plasma, thereby, converting their kinetic energy to high temperature heat which can be continuously removed by a coolant stream and converted in part to electricity in a conventional power turbine. Because of the primary goal of achieving minimum radioactivity, to date Brookhaven blanket concepts have been restricted to the use of some form of solid lithium, with inert gas-cooling and in some design cases, water-cooling of the shell structure. Aluminum and graphite have been identified as very promising structural materials for fusion blankets, and conceptual designs based on these materials have been made. Depending on the thermal loading on the ''first'' wall which surrounds the plasma as well as blanket design, heat transfer problems may be noticeably different in gas-cooled solid blankets. Approaches to solution of heat removal problems as well as explanation of: (a) the after-heat problems in blankets; (b) tritium breeding in solids; and (c) materials selection for radiation shields relative to the minimum activity blanket efforts at Brookhaven are discussed.

  5. Application of Hastelloy X in gas-cooled reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

    1976-10-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given.

  6. Cost estimate guidelines for advanced nuclear power technologies

    Energy Technology Data Exchange (ETDEWEB)

    Delene, J.G.; Hudson, C.R. II

    1993-05-01

    Several advanced power plant concepts are currently under development. These include the Modular High Temperature Gas Cooled Reactors, the Advanced Liquid Metal Reactor and the Advanced Light Water Reactors. One measure of the attractiveness of a new concept is its cost. Invariably, the cost of a new type of power plant will be compared with other alternative forms of electrical generation. This report provides a common starting point, whereby the cost estimates for the various power plants to be considered are developed with common assumptions and ground rules. Comparisons can then be made on a consistent basis. This is the second update of these cost estimate guidelines. Changes have been made to make the guidelines more current (January 1, 1992) and in response to suggestions made as a result of the use of the previous report. The principal changes are that the reference site has been changed from a generic Northeast (Middletown) site to a more central site (EPRI`s East/West Central site) and that reference bulk commodity prices and labor productivity rates have been added. This report is designed to provide a framework for the preparation and reporting of costs. The cost estimates will consist of the overnight construction cost, the total plant capital cost, the operation and maintenance costs, the fuel costs, decommissioning costs and the power production or busbar generation cost.

  7. High-resolution mid-infrared spectroscopy of buffer-gas-cooled methyltrioxorhenium molecules

    CERN Document Server

    Tokunaga, Sean; Tarbutt, M; Darquié, B

    2016-01-01

    We demonstrate cryogenic buffer-gas cooling of gas-phase methyltrioxorhenium (MTO). This molecule is closely related to chiral organometallic molecules where the parity-violating energy differences between enantiomers may be measurable. The molecules are produced with a rotational temperature of approximately 6~K by laser ablation of an MTO pellet inside a cryogenic helium buffer gas cell. Facilitated by the low temperature, we demonstrate absorption spectroscopy of the 10.2~$\\mu$m antisymmetric Re=O stretching mode of MTO with a resolution of 8~MHz and a frequency accuracy of 30~MHz. We partially resolve the hyperfine structure and measure the nuclear quadrupole coupling of the excited vibrational state.

  8. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  9. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  10. Evaluation of Gas-Cooled Pressurized Phosphoric Acid Fuel Cells for Electric Utility Power Generation

    Science.gov (United States)

    Faroque, M.

    1983-01-01

    Gas cooling is a more reliable, less expensive and a more simple alternative to conventional liquid cooling for heat removal from the phosphoric acid fuel cell (PAFC). The feasibility of gas-cooling was already demonstrated in atmospheric pressure stacks. Theoretical and experimental investigations of gas-cooling for pressurized PAFC are presented. Two approaches to gas cooling, Distributed Gas-Cooling (DIGAS) and Separated Gas-Cooling (SGC) were considered, and a theoretical comparison on the basis of cell performance indicated SGC to be superior to DIGAS. The feasibility of SGC was experimentally demonstrated by operating a 45-cell stack for 700 hours at pressure, and determining thermal response and the effect of other related parameters.

  11. Evaluation of gas cooling for pressurized phosphoric acid fuel cell stacks

    Science.gov (United States)

    Farooque, M.; Skok, A. J.; Maru, H. C.; Kothmann, R. E.; Harry, R. W.

    1983-01-01

    Gas cooling is a more reliable, less expensive and a more simple alternative to conventional liquid cooling for heat removal from the phosphoric acid fuel cell (PAFC). The feasibility of gas cooling has already been demonstrated in atmospheric pressure stacks. This paper presents theoretical and experimental investigation of gas cooling for pressurized PAFC. Two approaches to gas cooling, Distributed Gas Cooling (DIGAS) and Separated Gas Cooling (SGC) were considered, and a theoretical comparison on the basis of cell performance indicated SGC to be superior to DIGAS. The feasibility of SGC was experimentally demonstrated by operating a 45-cell stack for 700 hours at pressure, and determining thermal response and the effect of other related parameters.

  12. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    Energy Technology Data Exchange (ETDEWEB)

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

  13. Nuclear integrated database and design advancement system

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-01-01

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

  14. Intense Atomic and Molecular Beams via Neon Buffer Gas Cooling

    CERN Document Server

    Patterson, David; Doyle, John M

    2008-01-01

    We realize a continuous guided beam of cold deuterated ammonia with a flux of 3e11 ND3 molecules/s and a continuous free-space beam of cold potassium with a flux of 1e16 K atoms/s. A novel feature of the buffer gas source used to produce these beams is cold neon, which, due to intermediate Knudsen number beam dynamics, produces a forward velocity and low-energy tail that is comparable to much colder helium-based sources. We expect this source to be trivially generalizable to a very wide range of atomic and molecular species with significant vapor pressure below 1000 K. This source has properties that make it a good starting point for laser cooling of molecules or atoms, cold collision studies, trapping, or nonlinear optics in buffer-gas-cooled atomic or molecular gases.

  15. 10MW高温气冷实验堆的堆体结构特点%Features of Reactor Structure Design for 10MW High Temperature Gas-Cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    刘俊杰; 王敏稚; 张征明; 张振声; 何树延

    2001-01-01

    模块式高温气冷堆是当今世界上公认的先进反应堆堆型之一。固有安全性是它的最突出的优点。本文对 10 MW高温气冷堆的堆体布置进行了详细描述,并对 10MW高温气冷堆的结构设计特点进行了分析。根据 10 MW高温气冷堆的特点,本文对该堆的固有安全性、制造工艺等方面的优点进行了论述。%As the world-wide accepted advanced nuclear reactor,the most prominent characteristic of High-temperature Gas-cooled Modular Reactor is its inherent safety.The core arrangement of the 10MW High-temperature Gas-cooled Reactor,which is now under construction,is described in detail.The features of its structure design are analyzed,and based on these features,the advantages of inherent safety and manufacture are also discussed.

  16. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  17. Recent advances in understanding nuclear size and shape.

    Science.gov (United States)

    Mukherjee, Richik N; Chen, Pan; Levy, Daniel L

    2016-04-25

    Size and shape are important aspects of nuclear structure. While normal cells maintain nuclear size within a defined range, altered nuclear size and shape are associated with a variety of diseases. It is unknown if altered nuclear morphology contributes to pathology, and answering this question requires a better understanding of the mechanisms that control nuclear size and shape. In this review, we discuss recent advances in our understanding of the mechanisms that regulate nuclear morphology, focusing on nucleocytoplasmic transport, nuclear lamins, the endoplasmic reticulum, the cell cycle, and potential links between nuclear size and size regulation of other organelles. We then discuss the functional significance of nuclear morphology in the context of early embryonic development. Looking toward the future, we review new experimental approaches that promise to provide new insights into mechanisms of nuclear size control, in particular microfluidic-based technologies, and discuss how altered nuclear morphology might impact chromatin organization and physiology of diseased cells.

  18. Role of gas cooling in tomorrow`s energy services industry

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, P.J.

    1997-04-01

    This article discusses the marketing approach and opportunities for suppliers and manufacturers of gas cooling equipment to partner with energy service companies (ESCOs). The author`s viewpoint is that in educating and partnering with ESCOs the gas cooling industry enables their technology to reach its potential in the projects that the ESCOs develop.

  19. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  20. Development of GAMMA Code and Evaluation for a Very High Temperature gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Chang H; Lim, H.S.; Kim, E.S.; NO, H.C.

    2007-06-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. This paper will also include what improvements will be made in the Gamma code for the VHTR.

  1. Primordial gas cooling behind shock waves in merging halos

    CERN Document Server

    Vasiliev, E O; Shchekinov, Yu.A.

    2006-01-01

    We investigate thermal regime of the baryons behind shock waves arising in the process of virialization of dark matter halos. We find a fraction of the shocked gas cooled by radiation of HD molecules down to the temperature of the cosmic microwave background (CMB): this fraction increases sharply from about $f_{\\rm c}\\sim 10^{-3}$ for dark halos of $M=5\\times 10^7\\msun$ to $\\sim 0.1$ for halos with $M=10^8\\msun$ at $z=10$. We show, however, that further increase of the mass does not lead to a significant growth of $f_{\\rm c}$ -- the assymptotic value for $M\\gg 10^8\\msun$ is of 0.2. We estimate star formation rate associated with such shock waves, and show that it can be a small but not negligible fraction of the star formation connected with cooling by HI and H$_2$. We argue that extremely metal-poor low-mass stars in the Milky Way may have been formed from primordial gas behind such shocks.

  2. A 50-100 kWe gas-cooled reactor for use on Mars.

    Energy Technology Data Exchange (ETDEWEB)

    Peters, Curtis D. (.)

    2006-04-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  3. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600.

  4. Gas-Cooled Fast Reactor: A Historical Overview and Future Outlook

    Directory of Open Access Journals (Sweden)

    W. F. G. van Rooijen

    2009-01-01

    Full Text Available A review is given of developments in the area of Gas-Cooled Fast Reactors (GCFR in the period from roughly 1960 until 1980. During that period, the GCFR concept was expected to increase the breeding gain, the thermal efficiency of a nuclear power plant, and alleviate some of the problems associated with liquid metal coolants. During this period, the GCFR concept was found to be more challenging than liquid-metal-cooled reactors, and none were ever constructed. In the second part of the paper, we provide an overview of the investigations on GCFR since the year 2000, when the Generation IV Initiative rekindled interest in this reactor type. The new GCFR concepts focus primarily on sustainable nuclear power, with very efficient resource use, minimum waste, and a very strong focus on (passive safety. An overview is presented of the main design characteristics of these Gen IV GCFRs, and a literature list is provided to guide the interested reader towards more detailed publications.

  5. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  6. Effect of Target Configuration on the Neutronic Performance of the Gas-Cooled ADS

    CERN Document Server

    Biss, K; Shetty, N; Nabbi, R

    2013-01-01

    With the utilization of nuclear energy transuranic elements like Pu, Am and Cm are produced causing high, long term radioactivity and radio toxicity, respectively. To reduce the radiological impact on the environment and to the repository Partitioning and Transmutation is considered as an efficient way. In this respect comprehensive research works are performed at different research institutes worldwide. The results show that the transmutation of TRU is achieved with fast neutrons due to the higher fission probability. Based on Accelerator Driven Systems (ADS) those neutrons are used in a particular system, in which mainly liquid metal eutectic (lead bismuth) is used as coolant. The neutronic performance of an ADS system based on gas cooling was studied in this work by using the simulation tool MCNPX. The usage of the Monte-Carlo method in MCNPX allows the simulation of the physical processes in a 3D-model of the core. In dependence of the spallation target material and design several parameters like the mult...

  7. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  8. A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Hangbok Choi

    2013-01-01

    Full Text Available In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2, which is a compact gas-cooled fast reactor (GFR. The EM2 augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2 core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

  9. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  10. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    Science.gov (United States)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  11. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    Directory of Open Access Journals (Sweden)

    Shutang Zhu

    2008-01-01

    Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

  12. Current design efforts for the gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States)]. e-mail: Kevan.Weaver@inl.gov

    2005-07-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFC I) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

  13. Development of the Advanced Nuclear Safety Information Management (ANSIM) System

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Jae Min; Ko, Young Cheol; Song, Tai Gil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Korea has become a technically independent nuclear country and has grown into an exporter of nuclear technologies. Thus, nuclear facilities are increasing in significance at KAERI (Korea Atomic Energy Research Institute), and it is time to address the nuclear safety. The importance of nuclear safety cannot be overemphasized. Therefore, a management system is needed urgently to manage the safety of nuclear facilities and to enhance the efficiency of nuclear information. We have established ISP (Information Strategy Planning) for the Integrated Information System of nuclear facility and safety management. The purpose of this paper is to develop a management system for nuclear safety. Therefore, we developed the Advanced Nuclear Safety Information Management system (hereinafter referred to as the 'ANSIM system'). The ANSIM system has been designed and implemented to computerize nuclear safety information for standardization, integration, and sharing in real-time. Figure 1 shows the main home page of the ANSIM system. In this paper, we describe the design requirements, contents, configurations, and utilizations of the ANSIM system

  14. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

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

  15. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program: Topical report I, selection of candidate alloys. Volume 3. Selection of surface coating/substrate systems for screening creep and structural stability studies

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-20

    Considering the high temperature, low O/sub 2/, high C environment of operation in the Very High Temperature Reactor (VHTR) Systems, the utilization of coatings is envisaged to hold potential for extending component lifetimes through the formation of stable and continuous oxide films with enhanced resistance to C diffusion. A survey of the current state of technology for high temperature coatings has been performed. The usefulness of these coatings on the Mo, Ni, and Fe base alloys is discussed. Specifically, no coating substitute was identified for TZM other than the well known W-3 (pack silicide) and Al/sub 2/O/sub 3/ forming coatings were recommended for the Fe and Ni base structural materials. Recommendations as to coating types and processng have been made based on the predicted VHTR component size, shape, base metal and operational environment. Four tests designed to evaluate the effects of selected combinations of coatings and substrate matrices are recommended for consideration.

  16. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  17. Survey of advanced nuclear technologies for potential applications of sonoprocessing.

    Science.gov (United States)

    Rubio, Floren; Blandford, Edward D; Bond, Leonard J

    2016-09-01

    Ultrasonics has been used in many industrial applications for both sensing at low power and processing at higher power. Generally, the high power applications fall within the categories of liquid stream degassing, impurity separation, and sonochemical enhancement of chemical processes. Examples of such industrial applications include metal production, food processing, chemical production, and pharmaceutical production. There are many nuclear process streams that have similar physical and chemical processes to those applications listed above. These nuclear processes could potentially benefit from the use of high-power ultrasonics. There are also potential benefits to applying these techniques in advanced nuclear fuel cycle processes, and these benefits have not been fully investigated. Currently the dominant use of ultrasonic technology in the nuclear industry has been using low power ultrasonics for non-destructive testing/evaluation (NDT/NDE), where it is primarily used for inspections and for characterizing material degradation. Because there has been very little consideration given to how sonoprocessing can potentially improve efficiency and add value to important process streams throughout the nuclear fuel cycle, there are numerous opportunities for improvement in current and future nuclear technologies. In this paper, the relevant fundamental theory underlying sonoprocessing is highlighted, and some potential applications to advanced nuclear technologies throughout the nuclear fuel cycle are discussed.

  18. Human factors aspects of advanced instrumentation in the nuclear industry

    Energy Technology Data Exchange (ETDEWEB)

    Carter, R.J.

    1989-01-01

    An important consideration in regards to the use of advanced instrumentation in the nuclear industry is the interface between the instrumentation system and the human. A survey, oriented towards identifying the human factors aspects of digital instrumentation, was conducted at a number of United States (US) and Canadian nuclear vendors and utilities. Human factors issues, subsumed under the categories of computer-generated displays, controls, organizational support, training, and related topics were identified. 20 refs., 2 tabs.

  19. Importance of Advanced Planning of Manufacturing for Nuclear Industry

    Directory of Open Access Journals (Sweden)

    Shykinov Nick

    2016-06-01

    Full Text Available In the context of energy demands by growing economies, climate changes, fossil fuel pricing volatility, and improved safety and performance of nuclear power plants, many countries express interest in expanding or acquiring nuclear power capacity. In the light of the increased interest in expanding nuclear power the supply chain for nuclear power projects has received more attention in recent years. The importance of the advanced planning of procurement and manufacturing of components of nuclear facilities is critical for these projects. Many of these components are often referred to as long-lead items. They may be equipment, products and systems that are identified to have a delivery time long enough to affect directly the overall timing of a project. In order to avoid negatively affecting the project schedule, these items may need to be sourced out or manufactured years before the beginning of the project. For nuclear facilities, long-lead items include physical components such as large pressure vessels, instrumentation and controls. They may also mean programs and management systems important to the safety of the facility. Authorized nuclear operator training, site evaluation programs, and procurement are some of the examples. The nuclear power industry must often meet very demanding construction and commissioning timelines, and proper advanced planning of the long-lead items helps manage risks to project completion time. For nuclear components there are regulatory and licensing considerations that need to be considered. A national nuclear regulator must be involved early to ensure the components will meet the national legal regulatory requirements. This paper will discuss timing considerations to address the regulatory compliance of nuclear long-lead items.

  20. Advanced Research Workshop on Preparedness for Nuclear and Radiological Threats

    CERN Document Server

    Diamond, David; Nuclear Threats and Security Challenges

    2015-01-01

    With the dissolution of the Soviet Union the nuclear threats facing the world are constantly evolving and have grown more complex since the end of the Cold War. The diversion of complete weapon systems or nuclear material to rogue nations and terrorist organizations has increased. The events of the past years have proved the necessity to reevaluate these threats on a level never before considered.  In recognition that no single country possesses all of the answers to the critical scientific, institutional and legal questions associated with combating nuclear and radiological terrorism, the NATO Advanced Research Workshop on “Preparedness for Nuclear and Radiological Threats” and this proceeding was structured to promote wide-ranging, multi-national exploration of critical technology needs and underlying scientific challenges to reducing the threat of nuclear/radiological terrorism; to illustrate through country-specific presentations how resulting technologies were used in national programs; and to outli...

  1. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  2. Gas Cooled, Natural Uranium, D20 Moderated Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahlberg, R.C.; Beasley, E.G.; DeBoer, T.K.; Evans, T.C.; Molino, D.F.; Rothwell, W.S.; Slivka, W.R.

    1956-08-01

    The attractiveness of a helium cooled, heavy water moderated, natural uranium central station power plant has been investigated. A fuel element has been devised which allows the D20 to be kept at a low pressure while the exit gas temperature is high. A preliminary cost analysis indicates that, using currently available materials, competitive nuclear power in foreign countries is possible.

  3. Integration of advanced nuclear materials separation processes

    Energy Technology Data Exchange (ETDEWEB)

    Jarvinen, G.D.; Worl, L.A.; Padilla, D.D.; Berg, J.M.; Neu, M.P.; Reilly, S.D.; Buelow, S.

    1998-12-31

    This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). This project has examined the fundamental chemistry of plutonium that affects the integration of hydrothermal technology into nuclear materials processing operations. Chemical reactions in high temperature water allow new avenues for waste treatment and radionuclide separation.Successful implementation of hydrothermal technology offers the potential to effective treat many types of radioactive waste, reduce the storage hazards and disposal costs, and minimize the generation of secondary waste streams. The focus has been on the chemistry of plutonium(VI) in solution with carbonate since these are expected to be important species in the effluent from hydrothermal oxidation of Pu-containing organic wastes. The authors investigated the structure, solubility, and stability of the key plutonium complexes. Installation and testing of flow and batch hydrothermal reactors in the Plutonium Facility was accomplished. Preliminary testing with Pu-contaminated organic solutions gave effluent solutions that readily met discard requirements. A new effort in FY 1998 will build on these promising initial results.

  4. Cost estimate guidelines for advanced nuclear power technologies

    Energy Technology Data Exchange (ETDEWEB)

    Hudson, C.R. II

    1987-07-01

    To make comparative assessments of competing technologies, consistent ground rules must be applied when developing cost estimates. This document provides a uniform set of assumptions, ground rules, and requirements that can be used in developing cost estimates for advanced nuclear power technologies.

  5. Cost estimate guidelines for advanced nuclear power technologies

    Energy Technology Data Exchange (ETDEWEB)

    Delene, J.G.; Hudson, C.R. II.

    1990-03-01

    To make comparative assessments of competing technologies, consistent ground rules must be applied when developing cost estimates. This document provides a uniform set of assumptions, ground rules, and requirements that can be used in developing cost estimates for advanced nuclear power technologies. 10 refs., 8 figs., 32 tabs.

  6. Cost estimate guidelines for advanced nuclear power technologies

    Energy Technology Data Exchange (ETDEWEB)

    Hudson, C.R. II

    1986-07-01

    To make comparative assessments of competing technologies, consistent ground rules must be applied when developing cost estimates. This document provides a uniform set of assumptions, ground rules, and requirements that can be used in developing cost estimates for advanced nuclear power technologies.

  7. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  8. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  9. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  10. ADVANCED CERAMIC MATERIALS FOR NEXT-GENERATION NUCLEAR APPLICATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.

    2010-09-29

    , and proliferation), the worldwide community is working to develop and deploy new nuclear energy systems and advanced fuel cycles. These new nuclear systems address the key challenges and include: (1) extracting the full energy value of the nuclear fuel; (2) creating waste solutions with improved long term safety; (3) minimizing the potential for the misuse of the technology and materials for weapons; (4) continually improving the safety of nuclear energy systems; and (5) keeping the cost of energy affordable.

  11. Advanced nuclear systems. Review study; Fortgeschrittene Nuklearsysteme. Review Study

    Energy Technology Data Exchange (ETDEWEB)

    Liebert, Wolfgang; Glaser, Alexander; Pistner, Christoph [Interdisziplinaere Arbeitsgruppe Naturwissenschaft, Technik und Sicherheit (IANUS), Darmstadt University of Technology, Hochschulstrasse 10, D-64289 Darmstadt (Germany); Baehr, Roland; Hahn, Lothar [Institute for applied ecology (Oeko-Institut), Elisabethenstrasse 55-57, D-64283 Darmstadt (Germany)

    1999-04-01

    The task of this review study is to from provide an overview of the developments in the field of the various advanced nuclear systems, and to create the basis for more comprehensive studies of technology assessment. In an overview the concepts for advanced nuclear systems pursued worldwide are subdivided into eight subgroups. A coarse examination raster (set pattern) is developed to enable a detailed examination of the selected systems. In addition to a focus on enhanced safety features, further aspects are also taken into consideration, like the lowering of the proliferation risk, the enhancement of the economic competitiveness of the facilities and new usage possibilities (for instance concerning the relaxation of the waste disposal problem or the usage of alternative fuels to uranium). The question about the expected time span for realization and the discussion about the obstacles on the way to a commercially usable reactor also play a substantial role as well as disposal requirements as far as they can be presently recognized. In the central chapter of this study, the documentation of the representatively selected concepts is evaluated as well as existing technology assessment studies and expert opinions. In a few cases where this appears to be necessary, according technical literature, further policy advisory reports, expert statements as well as other relevant sources are taken into account. Contradictions, different assessments and dissents in the literature as well as a few unsettled questions are thus indicated. The potential of advanced nuclear systems with respect to economical and societal as well as environmental objectives cannot exclusively be measured by the corresponding intrinsic or in comparison remarkable technical improvements. The acceptability of novel or improved systems in nuclear technology will have to be judged by their convincing solutions for the crucial questions of safety, nuclear waste and risk of proliferation of nuclear weapons

  12. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  13. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  14. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  15. Production of liquid fuels with a high-temperature gas-cooled reactor

    Science.gov (United States)

    Quade, R. N.; Vrable, D. L.; Green, L., Jr.

    An exploration is made of the technical, economic and environmental impact feasibility of integrating coal liquefaction methods directly and indirectly with a nuclear reactor source of process heat, with stress on the production of synthetic jet fuel. Production figures and operating costs are compared for indirect conventional and nuclear processes using Lurgi-Fischer-Tropsch technology with direct conventional and nuclear techniques employing the advanced SRC-II technology, and it is concluded that significant advantages in coal savings and environmental impact can be expected from nuclear reactor integration.

  16. Advanced nuclear reactor public opinion project. Interim report

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-11

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-02-01

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

  19. Advanced ceramic materials for next-generation nuclear applications

    Science.gov (United States)

    Marra, John

    2011-10-01

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

  20. Analysis of Fluid Flow and Heat Transfer Model for the Pebble Bed High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    S. Yamoah

    2012-06-01

    Full Text Available The pebble bed type high temperature gas cooled nuclear reactor is a promising option for next generation reactor technology and has the potential to provide high efficiency and cost effective electricity generation. The reactor unit heat transfer poses a challenge due to the complexity associated with the thermalflow design. Therefore to reliably simulate the flow and heat transport of the pebble bed modular reactor necessitates a heat transfer model that deals with radiation as well as thermal convection and conduction. In this study, a model with the capability to simulate fluid flow and heat transfer in the pebble bed modular reactor core has been developed. The developed model was implemented on a personal computer using FORTRAN 95 programming language. Several important fluid flow and heat transfer parameters have been examined: including the pressure drop over the reactor core, the heat transfer coefficient, the Nusselt number and the effective thermal conductivity of the fuel pebbles. Results obtained from the simulation experiments show a uniform pressure in the radial direction for a core to fuel element diameter (D/d ratio>20 and the heat transfer coefficient increases with increasing temperature and coolant mass flow rate. The model can adequately account for the flow and heat transfer phenomenon and the loss of pressure through friction in the pebble bed type high temperature nuclear reactor.

  1. Output Feedback Dissipation Control for the Power-Level of Modular High-Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2011-11-01

    Full Text Available Because of its strong inherent safety features and the high outlet temperature, the modular high temperature gas-cooled nuclear reactor (MHTGR is the chosen technology for a new generation of nuclear power plants. Such power plants are being considered for industrial applications with a wide range of power levels, thus power-level regulation is very important for their efficient and stable operation. Exploiting the large scale asymptotic closed-loop stability provided by nonlinear controllers, a nonlinear power-level regulator is presented in this paper that is based upon both the techniques of feedback dissipation and well-established backstepping. The virtue of this control strategy, i.e., the ability of globally asymptotic stabilization, is that it takes advantage of the inherent zero-state detectability property of the MHTGR dynamics. Moreover, this newly built power-level regulator is also robust towards modeling uncertainty in the control rod dynamics. If modeling uncertainty of the control rod dynamics is small enough to be omitted, then this control law can be simplified to a classical proportional feedback controller. The comparison of the control performance between the newly-built power controller and the simplified controller is also given through numerical study and theoretical analysis.

  2. Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, Mark D [ORNL; Pritchard, Megan L [ORNL

    2008-01-01

    The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the

  3. Comb-assisted cavity ring-down spectroscopy of a buffer-gas-cooled molecular beam.

    Science.gov (United States)

    Santamaria, Luigi; Sarno, Valentina Di; Natale, Paolo De; Rosa, Maurizio De; Inguscio, Massimo; Mosca, Simona; Ricciardi, Iolanda; Calonico, Davide; Levi, Filippo; Maddaloni, Pasquale

    2016-06-22

    We demonstrate continuous-wave cavity ring-down spectroscopy of a partially hydrodynamic molecular beam emerging from a buffer-gas-cooling source. Specifically, the (ν1 + ν3) vibrational overtone band of acetylene (C2H2) around 1.5 μm is accessed using a narrow-linewidth diode laser stabilized against a GPS-disciplined rubidium clock via an optical frequency comb synthesizer. As an example, the absolute frequency of the R(1) component is measured with a fractional accuracy of ∼1 × 10(-9). Our approach represents the first step towards the extension of more sophisticated cavity-enhanced interrogation schemes, including saturated absorption cavity ring-down or two-photon excitation, to buffer-gas-cooled molecular beams.

  4. Nuclear reactor design

    CERN Document Server

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Walter, C E

    1999-05-01

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

  6. Advanced Space Nuclear Reactors from Fiction to Reality

    Science.gov (United States)

    Popa-Simil, L.

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

  7. THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    David S. Duncan; Vondell J. Balls; Stephanie L. Austad

    2008-09-01

    The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

  8. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  9. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  10. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. Novel Applications of Buffer-Gas Cooling to Cold Atoms, Diatomic Molecules, and Large Molecules

    OpenAIRE

    Drayna, Garrett Korda

    2016-01-01

    Cold gases of atoms and molecules provide a system for the exploration of a diverse set of physical phenomena. For example, cold gasses of magnetically and electrically polar atoms and molecules are ideal systems for quantum simulation and quantum computation experiments, and cold gasses of large polar molecules allow for novel spectroscopic techniques. Buffer-gas cooling is a robust and widely applicable method for cooling atoms and molecules to temperatures of approximately 1 Kelvin. In thi...

  12. Novel Applications of Buffer-gas Cooling to Cold Atoms, Diatomic Molecules, and Large Molecules

    Science.gov (United States)

    Drayna, Garrett Korda

    Cold gases of atoms and molecules provide a system for the exploration of a diverse set of physical phenomena. For example, cold gasses of magnetically and electrically polar atoms and molecules are ideal systems for quantum simulation and quantum computation experiments, and cold gasses of large polar molecules allow for novel spectroscopic techniques. Buffer-gas cooling is a robust and widely applicable method for cooling atoms and molecules to temperatures of approximately 1 Kelvin. In this thesis, I present novel applications of buffer-gas cooling to obtaining gases of trapped, ultracold atoms and diatomic molecules, as well as the study of the cooling of large organic molecules. In the first experiment of this thesis, a buffer-gas beam source of atoms is used to directly load a magneto-optical trap. Due to the versatility of the buffer-gas beam source, we obtain trapped, sub-milliKelvin gases of four different lanthanide species using the same experimental apparatus. In the second experiment of this thesis, a buffer-gas beam is used as the initial stage of an experiment to directly laser cool and magneto-optically trap the diatomic molecule CaF. In the third experiment of this thesis, buffer-gas cooling is used to study the cooling of the conformational state of large organic molecules. We directly observe conformational relaxation of gas-phase 1,2-propanediol due to cold collisions with helium gas. Lastly, I present preliminary results on a variety of novel applications of buffer-gas cooling, such as mixture analysis, separation of chiral mixtures, the measurement of parity-violation in chiral molecules, and the cooling and spectroscopy of highly unstable reaction intermediates.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  14. Bibliography of Connecticut Advanced Nuclear Engineering Laboratory reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-01

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

  15. Bibliography of Connecticut Advanced Nuclear Engineering Laboratory reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-01

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

  16. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Science.gov (United States)

    2010-01-01

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

  17. [USA/FRG cooperation in gas-cooled reactor development]. Foreign trip report, June 24--July 2, 1988

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Jr, J E

    1988-07-26

    Reviews were conducted at Kernforschungsanlage (KFA) Juelich of the US and Federal Republic of Germany (FRG) high-temperature gas-cooled reactor (HTGR) programs under the US/FRG Umbrella Agreement, with emphasis on those technology development areas where cooperation is ongoing and planned. Specific subprogram areas are safety; materials; fuels, fission products, and graphite; and Arbeitsgemeinschaft Versuchs-Reaktor (AVR). The purpose was to assess the status of the cooperation, reach agreement on any changes needed, and identify new areas of cooperation. Overall, the agreement has been both effective and beneficial. Ongoing activities complement and support US technology development plans. Discussions were held in the United Kingdom (UK) at the Risley Nuclear Power Development Laboratory regarding a potential graphite technology exchange program between the US Department of Energy and the UK Atomic Energy Authority. A draft agreement was reviewed and appeared to be satisfactory to both parties and ready for signature. A summary of potential areas of activity in the exchange had been prepared by US representatives and was discussed and found to be acceptable to UK representatives.

  18. An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-02-01

    Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

  19. Advances in Nuclear Magnetic Resonance for Drug Discovery

    Science.gov (United States)

    Powers, Robert

    2010-01-01

    Background Drug discovery is a complex and unpredictable endeavor with a high failure rate. Current trends in the pharmaceutical industry have exasperated these challenges and are contributing to the dramatic decline in productivity observed over the last decade. The industrialization of science by forcing the drug discovery process to adhere to assembly-line protocols is imposing unnecessary restrictions, such as short project time-lines. Recent advances in nuclear magnetic resonance are responding to these self-imposed limitations and are providing opportunities to increase the success rate of drug discovery. Objective/Method A review of recent advancements in NMR technology that have the potential of significantly impacting and benefiting the drug discovery process will be presented. These include fast NMR data collection protocols and high-throughput protein structure determination, rapid protein-ligand co-structure determination, lead discovery using fragment-based NMR affinity screens, NMR metabolomics to monitor in vivo efficacy and toxicity for lead compounds, and the identification of new therapeutic targets through the functional annotation of proteins by FAST-NMR. Conclusion NMR is a critical component of the drug discovery process, where the versatility of the technique enables it to continually expand and evolve its role. NMR is expected to maintain this growth over the next decade with advancements in automation, speed of structure calculation, in-cell imaging techniques, and the expansion of NMR amenable targets. PMID:20333269

  20. Dedicated nuclear facilities for electrolytic hydrogen production

    Science.gov (United States)

    Foh, S. E.; Escher, W. J. D.; Donakowski, T. D.

    1979-01-01

    An advanced technology, fully dedicated nuclear-electrolytic hydrogen production facility is presented. This plant will produce hydrogen and oxygen only and no electrical power will be generated for off-plant use. The conceptual design was based on hydrogen production to fill a pipeline at 1000 psi and a 3000 MW nuclear base, and the base-line facility nuclear-to-shaftpower and shaftpower-to-electricity subsystems, the water treatment subsystem, electricity-to-hydrogen subsystem, hydrogen compression, efficiency, and hydrogen production cost are discussed. The final conceptual design integrates a 3000 MWth high-temperature gas-cooled reactor operating at 980 C helium reactor-out temperature, direct dc electricity generation via acyclic generators, and high-current density, high-pressure electrolyzers based on the solid polymer electrolyte approach. All subsystems are close-coupled and optimally interfaced and pipeline hydrogen is produced at 1000 psi. Hydrogen costs were about half of the conventional nuclear electrolysis process.

  1. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

    IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  2. HFE safety reviews of advanced nuclear power plant control rooms

    Science.gov (United States)

    Ohara, John

    1994-01-01

    Advanced control rooms (ACR's) will utilize human-system interface (HSI) technologies that may have significant implications for plant safety in that they will affect the operator's overall role and means of interacting with the system. The Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) aspects of HSI's to ensure that they are designed to good HFE principles and support performance and reliability in order to protect public health and safety. However, the only available NRC guidance was developed more than ten years ago, and does not adequately address the human performance issues and technology changes associated with ACR's. Accordingly, a new approach to ACR safety reviews was developed based upon the concept of 'convergent validity'. This approach to ACR safety reviews is described.

  3. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  4. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  5. Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-11-01

    Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

  6. Advanced nuclear power plants for the 21st century

    Energy Technology Data Exchange (ETDEWEB)

    Graham, J. [BNFL, Inc. (United States)

    1996-10-01

    This paper examines, following four issues: capacity; the closure of the fuel cycle; deregulation; and the need to maintain the development of the advanced systems. Demand is a governing parameter: if one doesn`t need the power then there is no need to increase generating capacity. However, there is no question but that the population is growing. All predictions are that new generating capacity will be needed -- the questions are when and how? Until the various issues involved in deregulation are played through it is not clear what form markets will take for the longer term or how investment in large-capital-cost facilities will fit into the financial structure. Deregulation needs the time to throw light on these matters and to gain some experience in the various financial options. The lack of closure of the fuel cycle is both a cost and public perception issue. The US program, as a result of a cold-war paranoia against recycling the partially used fuel, is based upon the final disposal of useful supplies of energy. However, the program itself is plagued with poor management, delays, and uncertainties that are due, in no small measure, to half-uttered thoughts by all concerned, that this is the wrong policy. Current efforts to rethink the policy, and its implementing projects, are welcome. Finally, if it is important to keep design options for advanced nuclear power plants open for the future, then it necessary to maintain valid research and development programs for those designs. Current US policy is damaging to a number of the more advanced options. This paper discusses the candidate systems: LWR, ALMR, HTGR, and CANDU systems for the special contributions they may each provide in an ideal electrical generating industry of the mid-twenty-first century, and makes suggestions for the future. (J.P.N.)

  7. A review on the development of the advanced fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Lee, Yung Woo; Sohn, Dong Sung; Yang, Myung Seung; Bae, Kee Kwang; Nah, Sang Hoh; Kim, Han Soo; Kim, Bong Koo; Song, Keun Woo; Kim, See Hyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    In this state-of art report, the development status of the advanced nuclear fuel was investigated. The current fabrication technology for coated particle fuel and non-oxide fuel such as sol-gel technology, coating technology, and carbothermic reduction reaction has also been examined. In the view point of inherent safety and efficiency in the operation of power plant, the coated particle fuel will keep going on its reputation as nuclear fuel for a high temperature gas cooled reactor, and the nitride fuel is very prospective for the next liquid metal fast breeder reactor. 43 figs., 17 tabs., 96 refs. (Author).

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

    Science.gov (United States)

    Carlsen, Robert W.

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

  9. Advancement of safeguards inspection technology for CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Sung; Park, W. S.; Cha, H. R.; Ham, Y. S.; Lee, Y. G.; Kim, K. P.; Hong, Y. D

    1999-04-01

    The objectives of this project are to develop both inspection technology and safeguards instruments, related to CANDU safeguards inspection, through international cooperation, so that those outcomes are to be applied in field inspections of national safeguards. Furthermore, those could contribute to the improvement of verification correctness of IAEA inspections. Considering the level of national inspection technology, it looked not possible to perform national inspections without the joint use of containment and surveillance equipment conjunction with the IAEA. In this connection, basic studies for the successful implementation of national inspections was performed, optimal structure of safeguards inspection was attained, and advancement of safeguards inspection technology was forwarded. The successful implementation of this project contributed to both the improvement of inspection technology on CANDU reactors and the implementation of national inspection to be performed according to the legal framework. In addition, it would be an opportunity to improve the ability of negotiating in equal shares in relation to the IAEA on the occasion of discussing or negotiating the safeguards issues concerned. Now that the national safeguards technology for CANDU reactors was developed, the safeguards criteria, procedure and instruments as to the other item facilities and fabrication facilities should be developed for the perfection of national inspections. It would be desirable that the recommendations proposed and concreted in this study, so as to both cope with the strengthened international safeguards and detect the undeclared nuclear activities, could be applied to national safeguards scheme. (author)

  10. Meeting nuclear data needs for advanced reactor system

    OpenAIRE

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

    2014-01-01

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

  11. Safety and licensing of MHTGR (Modular High Temperature Gas Cooled Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Silady, F.A.; Millunzi, A.C.; Kelley, A.P. Jr.; Cunliffe, J.

    1987-07-01

    The Modular High Temperature Gas Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant boundary. The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error.

  12. Simultaneous approach for simulation of a high-temperature gas-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Yang CHEN; Jiang-hong YOU; Zhi-jiang SHAO; Ke-xin WANG; Ji-xin QIAN

    2011-01-01

    The simulation of a high-temperature gas-cooled reactor pebble-bed module (HTR-PM) plant is discussed.This lumped parameter model has the form of a set differential algebraic equations (DAEs) that include stiff equations to model point neutron kinetics.The nested approach is the most common method to solve DAE,but this approach is very expensive and time-consuming due to inner iterations.This paper deals with an alternative approach in which a simultaneous solution method is used.The DAEs are discretized over a time horizon using collocation on finite elements,and Radau collocation points are applied.The resulting nonlinear algebraic equations can be solved by existing solvers.The discrete algorithm is discussed in detail; both accuracy and stability issues are considered.Finally,the simulation results are presented to validate the efficiency and accuracy of the simultaneous approach that takes much less time than the nested one.

  13. Physics with Cold Molecules Using Buffer Gas Cooling: Precision Measurement, Collisions, and Laser Cooling

    Science.gov (United States)

    Hutzler, Nicholas R.; Doyle, John M.

    2014-06-01

    Cryogenic buffer gas cooled beams and cells can be used to study many species, from atoms and polar molecules to biomolecules. We report on recent applications of this technique to improve the limit on the electron electric dipole moment [1], load polar molecules into a magnetic trap through optical pumping [2], perform chirally sensitive microwave spectroscopy on polyatomic molecules [3], progress towards magneto-optical trapping of polar molecules [4], and studies of atom-molecule sticking [5]. [1] The ACME Collaboration: J. Baron et al., Science 343, p. 269 (2014) [2] B. Hemmerling et al., arXiv:1310.2669, to appear in Phys. Rev. Lett. [3] D. Patterson, M. Schnell, & J. M. Doyle, Nature 497, p. 475 (2013) [4] H. Lu et al., arXiv:1310.3239, to appear in New. J. Phys. [5] J. Piskorski et al., under preparation

  14. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  15. Sensitivity and resolution in frequency comb spectroscopy of buffer gas cooled polyatomic molecules

    Science.gov (United States)

    Changala, P. Bryan; Spaun, Ben; Patterson, David; Doyle, John M.; Ye, Jun

    2016-12-01

    We discuss the use of cavity-enhanced direct frequency comb spectroscopy in the mid-infrared region with buffer gas cooling of polyatomic molecules for high-precision rovibrational absorption spectroscopy. A frequency comb coupled to an optical enhancement cavity allows us to collect high-resolution, broad-bandwidth infrared spectra of translationally and rotationally cold (10-20 K) gas-phase molecules with high absorption sensitivity and fast acquisition times. The design and performance of the combined apparatus are discussed in detail. Recorded rovibrational spectra in the CH stretching region of several organic molecules, including vinyl bromide (CH_2CHBr), adamantane (C_{10}H_{16}), and diamantane (C_{14}H_{20}) demonstrate the resolution and sensitivity of this technique, as well as the intrinsic challenges faced in extending the frontier of high-resolution spectroscopy to large complex molecules.

  16. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  17. Gas cooling in semi-analytic models and SPH simulations: are results consistent?

    CERN Document Server

    Saro, A; Borgani, S; Dolag, K

    2010-01-01

    We present a detailed comparison between the galaxy populations within a massive cluster, as predicted by hydrodynamical SPH simulations and by a semi-analytic model (SAM) of galaxy formation. Both models include gas cooling and a simple prescription of star formation, which consists in transforming instantaneously any cold gas available into stars, while neglecting any source of energy feedback. We find that, in general, galaxy populations from SAMs and SPH have similar statistical properties, in agreement with previous studies. However, when comparing galaxies on an object-by-object basis, we find a number of interesting differences: a) the star formation histories of the brightest cluster galaxies (BCGs) from SAM and SPH models differ significantly, with the SPH BCG exhibiting a lower level of star formation activity at low redshift, and a more intense and shorter initial burst of star formation with respect to its SAM counterpart; b) while all stars associated with the BCG were formed in its progenitors i...

  18. Contributions to the neutronic analysis of a gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reyes-Ramirez, Ricardo, E-mail: ricarera@yahoo.com.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reinking-Cejudo, Arturo G., E-mail: reinking@servidor.unam.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico)

    2011-06-15

    Highlights: > Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. > Fuel lattice and core criticality calculations were done. > A higher Doppler coefficient than coolant density coefficient. > Zirconium carbide is a better reflector than silicon carbide. > Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained

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

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

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

  20. Recent advances in maize nuclear proteomic studies reveal histone modifications.

    Science.gov (United States)

    Casati, Paula

    2012-01-01

    The nucleus of eukaryotic organisms is highly dynamic and complex, containing different types of macromolecules including DNA, RNA, and a wide range of proteins. Novel proteomic applications have led to a better overall determination of nucleus protein content. Although nuclear plant proteomics is only at the initial phase, several studies have been reported and are summarized in this review using different plants species, such as Arabidopsis thaliana, rice, cowpea, onion, garden cress, and barrel clover. These include the description of the total nuclear or phospho-proteome (i.e., Arabidopsis, cowpea, onion), or the analysis of the differential nuclear proteome under different growth environments (i.e., Arabidopsis, rice, cowpea, onion, garden cress, and barrel clover). However, only few reports exist on the analysis of the maize nuclear proteome or its changes under various conditions. This review will present recent data on the study of the nuclear maize proteome, including the analysis of changes in posttranslational modifications in histone proteins.

  1. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  2. Human factors design review guidelines for advanced nuclear control room technologies

    Energy Technology Data Exchange (ETDEWEB)

    O' Hara, J.; Brown, W. (Brookhaven National Lab., Upton, NY (United States)); Granda, T.; Baker, C. (Carlow Associates, Inc., Fairfax, VA (United States))

    1991-01-01

    Advanced control rooms (ACRs) for future nuclear power plants are being designed utilizing computer-based technologies. The US Nuclear Regulatory Commission reviews the human engineering aspects of such control rooms to ensure that they are designed to good human factors engineering principles and that operator performance and reliability are appropriately supported in order to protect public health and safety. This paper describes the rationale, general approach, and initial development of an NRC Advanced Control Room Design Review Guideline. 20 refs., 1 fig.

  3. NATO Advanced Study Institute on New Aspects of Nuclear Dynamics

    CERN Document Server

    Huberts, P

    1989-01-01

    The 1988 Summer School on New Aspects of Nuclear Dynamics took place in the style that by now has become a tradition: a series of lectures by well known scientists on modem topics of nuclear physics, where special emphasis is placed on the didactic aspects of the lectures. In the past few years, we have witnessed a rapid evolution of the field of nuclear physics towards novel directions of research. This development is accompanied by the construction of some of the largest experimental facilities ever built for nuclear research. The subjects covered by the Summer School focussed on two main issues currently under active investigation and which will be pursued with the new facilities: the transition from nucleonic to quark degrees of freedom in the decription of nuclear reactions, and the behavior of nuclear matter as one approaches extreme densities and temperatures. These topics in many respects go beyond traditional nuclear physics and the speakers therefore also included high energy physicists. From the re...

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

    CERN Document Server

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

    2010-01-01

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

  5. The Advanced Controls Program at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Knee, H.E.; White, J.D.

    1990-01-01

    The Oak Ridge National Laboratory (ORNL), under sponsorship of the US Department of Energy (DOE), is conducting research that will lead to advanced, automated control of new liquid-metal-reactor (LMR) nuclear power plants. Although this program of research (entitled the Advanced Controls Program'') is focused on LMR technology, it will be capable of providing control design, test, and qualification capability for other advanced reactor designs (e.g., the advanced light water reactor (ALWR) and high temperature gas-cooled reactor (HTGR) designs), while also benefiting existing nuclear plants. The Program will also have applicability to complex, non-nuclear process control environments (e.g., petrochemical, aerospace, etc.). The Advanced Controls Program will support capabilities throughout the entire plant design life cycle, i.e., from the initial interactive first-principle dynamic model development for the process, systems, components, and instruments through advanced control room qualification. The current program involves five principal areas of research activities: (1) demonstrations of advanced control system designs, (2) development of an advanced controls design environment, (3) development of advanced control strategies, (4) research and development (R D) in human-system integration for advanced control system designs, and (5) testing and validation of advanced control system designs. Discussion of the research in these five areas forms the basis of this paper. Also included is a description of the research directions of the program. 8 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-04-15

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

  7. NATO Advanced Research Workshop on Preparedness for Nuclear and Radiological Threats

    CERN Document Server

    Diamond, David

    2015-01-01

    The nuclear crisis in Fukushima and growing threats of nuclear terrorism must serve as a wake-up call, prompting greater action to prepare ourselves for nuclear and radiological disasters. Our strategy to prepare for these threats is multi-layered and the events of these past years have proved the necessity to re-evaluate the national and international preparedness goals on a scale never before considered. The programme of NATO Advanced Research Workshop on “Preparedness for Nuclear and Radiological Threats” has been focused on science and technology challenges associated with our need to improve the national and international capacity and capability to prevent, protect against, mitigate the effects of, respond to, and recover from the nuclear and radiological disasters, including nuclear and radiological accident, terrorist attack by Improvised Nuclear Device (IND) or by “Dirty Bomb”-Radiological Dispersal Device (RDD), that pose the greatest risk to the national and international security and safety...

  8. Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

    Energy Technology Data Exchange (ETDEWEB)

    Lindemer, T.B.

    2002-09-05

    Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed.

  9. Efficiency of gas cooling and accretion at the disc-corona interface

    Science.gov (United States)

    Armillotta, L.; Fraternali, F.; Marinacci, F.

    2016-11-01

    In star-forming galaxies, stellar feedback can have a dual effect on the circumgalactic medium both suppressing and stimulating gas accretion. The trigger of gas accretion can be caused by disc material ejected into the halo in the form of fountain clouds and by its interaction with the surrounding hot corona. Indeed, at the disc-corona interface, the mixing between the cold/metal-rich disc gas (T ≲ 104 K) and the hot coronal gas (T ≳ 106 K) can dramatically reduce the cooling time of a portion of the corona and produce its condensation and accretion. We studied the interaction between fountain clouds and corona in different galactic environments through parsec-scale hydrodynamical simulations, including the presence of thermal conduction, a key mechanism that influences gas condensation. Our simulations showed that the coronal gas condensation strongly depends on the galactic environment, in particular it is less efficient for increasing virial temperature/mass of the haloes where galaxies reside and it is fully ineffective for objects with virial masses larger than 1013 M⊙. This result implies that the coronal gas cools down quickly in haloes with low-intermediate virial mass (Mvir ≲ 3 × 1012 M⊙) but the ability to cool the corona decreases going from late-type to early-type disc galaxies, potentially leading to the switching off of accretion and the quenching of star formation in massive systems.

  10. Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, M.L.

    1979-01-01

    This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.

  11. Procedure of Active Residual Heat Removal after Emergency Shutdown of High-Temperature-Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Xingtuan Yang

    2014-01-01

    Full Text Available After emergency shutdown of high-temperature-gas-cooled reactor, the residual heat of the reactor core should be removed. As the natural circulation process spends too long period of time to be utilized, an active residual heat removal procedure is needed, which makes use of steam generator and start-up loop. During this procedure, the structure of steam generator may suffer cold/heat shock because of the sudden load of coolant or hot helium at the first few minutes. Transient analysis was carried out based on a one-dimensional mathematical model for steam generator and steam pipe of start-up loop to achieve safety and reliability. The results show that steam generator should be discharged and precooled; otherwise, boiling will arise and introduce a cold shock to the boiling tubes and tube sheet when coolant began to circulate prior to the helium. Additionally, in avoiding heat shock caused by the sudden load of helium, the helium circulation should be restricted to start with an extreme low flow rate; meanwhile, the coolant of steam generator (water should have flow rate as large as possible. Finally, a four-step procedure with precooling process of steam generator was recommended; sensitive study for the main parameters was conducted.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-05-01

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

  13. Recent advances in maize nuclear proteomic studies reveal histone modifications

    Directory of Open Access Journals (Sweden)

    Paula eCasati

    2012-12-01

    Full Text Available The nucleus of eukaryotic organisms is highly dynamic and complex, containing different types of macromolecules including DNA, RNA, and a wide range of proteins. Novel proteomic applications have led to a better overall determination of nucleus protein content. Although nuclear plant proteomics is only at the initial phase, several studies have been reported and are summarized in this review using different plants species, such as Arabidopsis thaliana, rice, cowpea, onion, garden cress, and barrel clover. These include the description of the total nuclear or phospho-proteome (i.e. Arabidopsis, cowpea, onion, or the analysis of the differential nuclear proteome under different growth environments (i.e. Arabidopsis, rice, cowpea, onion, garden cress and barrel clover. However, only few reports exist on the analysis of the maize nuclear proteome or its changes under various conditions. This review will present recent data on the study of the nuclear maize proteome, including the analysis of changes in posttranslational modifications in histone proteins.

  14. Advanced methods for nuclear reactor gas laser coupling

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-06-01

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

  15. Recent Advances in Nuclear Powered Electric Propulsion for Space Exploration

    Science.gov (United States)

    Cassady, R. Joseph; Frisbee, Robert H.; Gilland, James H.; Houts, Michael G.; LaPointe, Michael R.; Maresse-Reading, Colleen M.; Oleson, Steven R.; Polk, James E.; Russell, Derrek; Sengupta, Anita

    2007-01-01

    Nuclear and radioisotope powered electric thrusters are being developed as primary in-space propulsion systems for potential future robotic and piloted space missions. Possible applications for high power nuclear electric propulsion include orbit raising and maneuvering of large space platforms, lunar and Mars cargo transport, asteroid rendezvous and sample return, and robotic and piloted planetary missions, while lower power radioisotope electric propulsion could significantly enhance or enable some future robotic deep space science missions. This paper provides an overview of recent U.S. high power electric thruster research programs, describing the operating principles, challenges, and status of each technology. Mission analysis is presented that compares the benefits and performance of each thruster type for high priority NASA missions. The status of space nuclear power systems for high power electric propulsion is presented. The paper concludes with a discussion of power and thruster development strategies for future radioisotope electric propulsion systems,

  16. Reference Operational Concepts for Advanced Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques Victor [Idaho National Lab. (INL), Idaho Falls, ID (United States); Farris, Ronald Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This report represents the culmination of a four-year research project that was part of the Instrumentation and Control and Human Machine Interface subprogram of the DOE Advanced Reactor Technologies program.

  17. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  18. The Shandong Shidao Bay 200 MWe High-Temperature Gas-Cooled Reactor Pebble-Bed Module (HTR-PM Demonstration Power Plant: An Engineering and Technological Innovation

    Directory of Open Access Journals (Sweden)

    Zuoyi Zhang

    2016-03-01

    Full Text Available After the first concrete was poured on December 9, 2012 at the Shidao Bay site in Rongcheng, Shandong Province, China, the construction of the reactor building for the world's first high-temperature gas-cooled reactor pebble-bed module (HTR-PM demonstration power plant was completed in June, 2015. Installation of the main equipment then began, and the power plant is currently progressing well toward connecting to the grid at the end of 2017. The thermal power of a single HTR-PM reactor module is 250 MWth, the helium temperatures at the reactor core inlet/outlet are 250/750 °C, and a steam of 13.25 MPa/567 °C is produced at the steam generator outlet. Two HTR-PM reactor modules are connected to a steam turbine to form a 210 MWe nuclear power plant. Due to China's industrial capability, we were able to overcome great difficulties, manufacture first-of-a-kind equipment, and realize series major technological innovations. We have achieved successful results in many aspects, including planning and implementing R&D, establishing an industrial partnership, manufacturing equipment, fuel production, licensing, site preparation, and balancing safety and economics; these obtained experiences may also be referenced by the global nuclear community.

  19. Development of advanced low alloy steel for nuclear RPV

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. C.; Shin, K. S.; Lee, S. H.; Lee, B. J. [Seoul National Univ., Seoul (Korea)

    2000-04-01

    Low carbon low alloy steels are used in nuclear power plants as pressure vessel, steam generator, etc. Nuclear pressure vessel material requires good combination of strength/ toughness, good weldability and high resistance to neutron irradiation and corrosion fatigue. For SA508III steels, most widely used in the production of nuclear power plant, attaining toughness is more difficult than strength. When taking into account the loss of toughness due to neutron irradiation, attaining as low transition temperature as possible prior to operation is a critical task in the production of nuclear pressure vessels. In the present study, we investigated detrimental microstructural features of SA508III steels to toughness, then alloy design directions to achieve improved mechanical properties were devised. The next step of alloy design was determined based on phase equilibrium thermodynamics and obtained results. Low carbon low alloy steels having low transition temperatures with enough strength and hardenability were developed. Microstructure and mechanical properties of HAZ of SA508III steels and alloy designed steels were investigated. 22 refs., 147 figs., 38 tabs. (Author)

  20. Progress of nuclear safety for symbiosis and sustainability advanced digital instrumentation, control and information systems for nuclear power plants

    CERN Document Server

    Yoshikawa, Hidekazu

    2014-01-01

    This book introduces advanced methods of computational and information systems allowing readers to better understand the state-of-the-art design and implementation technology needed to maintain and enhance the safe operation of nuclear power plants. The subjects dealt with in the book are (i) Full digital instrumentation and control systems and human?machine interface technologies (ii) Risk? monitoring methods for large and? complex? plants (iii) Condition monitors for plant components (iv) Virtual and augmented reality for nuclear power plants and (v) Software reliability verification and val

  1. High-Temperature Gas-Cooled Reactor Technology Development Program: Annual progress report for period ending December 31, 1987

    Energy Technology Data Exchange (ETDEWEB)

    Jones, J.E.,Jr.; Kasten, P.R.; Rittenhouse, P.L.; Sanders, J.P.

    1989-03-01

    The High-Temperature Gas-Cooled Reactor (HTGR) Program being carried out under the US Department of Energy (DOE) continues to emphasize the development of modular high-temperature gas-cooled reactors (MHTGRs) possessing a high degree of inherent safety. The emphasis at this time is to develop the preliminary design of the reference MHTGR and to develop the associated technology base and licensing infrastructure in support of future reactor deployment. A longer-term objective is to realize the full high-temperature potential of HTGRs in gas turbine and high-temperature, process-heat applications. This document summarizes the activities of the HTGR Technology Development Program for the period ending December 31, 1987.

  2. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  3. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.

    2002-10-01

    The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the

  4. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  5. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  6. Study on the properties of the fuel compact for High Temperature Gas-cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of); Lee, Young-woo; Cho, Moon-sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    High Temperature Gas-cooled Reactors (HTGR), one of the Gen-IV reactors, have been using the fuel element which is manufactured by the graphite matrix, surrounding Tristructural-isotropic (TRISO)-coated Uranium particles. Factors with these characteristics effecting on the matrix of fuel compact are chosen and their impacts on the properties are studied. The fuel elements are considered with two types of concepts for HTGR, which are the block type reactor and the pebble bed reactor. In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength with the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and the two kinds of candidate binder (Phenol and Polyvinyl butyral) were chosen and mixed with each other, formed and heated to measure mechanical properties. The objective of this research is to optimize the materials and composition of the mixture and the forming process by evaluating the mechanical properties before/after carbonization and heat treatment. From the mechanical test results, the mechanical properties of graphite pellets was related to the various conditions such as the contents and kinds of binder, the kinds of graphite and the heat treatments. In the result of the compressive strength and Vicker's hardness, the 10 wt% phenol binder added R+S graphite pellet was relatively higher mechanical properties than other pellets. The contents of Phenol binder, the kinds of graphite powder and the temperature of carbonization and heat treatment are considered important factors for the properties. To optimize the mechanical properties of fuel elements, the role of binders and the properties of graphites will be investigated as

  7. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  8. Evaluation of Thermoelectric Properties of Bite Alloys for the Optimization of Gas-Cooled Peltier Current Lead

    Science.gov (United States)

    Fujii, T.; Fukuda, S.; Kawahara, T.; Emoto, M.; Sugino, M.; Sun, J.; Hamabe, M.; Watanabe, H.; Yamaguchi, S.

    2010-04-01

    A direct current superconducting power transmission (DC-SCPT) test facility with 20-meter power cable has been developed in Chubu University. Heat leakage through current leads dominates in case of short-distance DC-SCPT. To solve this problem, we have developed special Peltier current lead (PCL) with bismuth telluride (BiTe) alloy is used as a Peltier material, and have studied a gas-cooled PCL as a novel type current lead for higher reduction of the heat leakage. In order to achieve high efficiency, highest thermoelectric figure of merit material is important to be used especially at low temperature conditions. In our work we measured the thermoelectric properties of BiTe samples, and optimized geometry of PCL by the numerical calculations. It was fount that the heat leakages through PCL and gas-cooled PCL are reduced to 55% and to 36% of the conventional copper current lead (CCL) at the self-cooling condition, correspondingly. PCL can reduce the total electric power consumption of the whole system. We also discuss the heat exchange ratio for the gas-cooled PCL, and extend this approach for further reduction of the heat leakage.

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

    Science.gov (United States)

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

    2010-09-01

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

  10. Advanced Elastic/Inelastic Nuclear Data Development Project

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, Frank; Chowdhury, Partha; Greife, Uwe; Fisher Hicks, Sally; Tsvetkov, Pavel; Rahn Vanhoy, Jeffrey; Hill, Tony; Kawano, Toshihiko; Slaughter, David

    2015-06-08

    The optical model is used to analyze the elastic and inelastic scattering of nucleons, deuterons, hellions, tritons, and alpha particles by the nuclei. Since this paper covers primarily neutron-nucleus scattering, the focus will be limited to only that interaction. For the sake of this model, the nucleus is described as a blob of nuclear matter with properties based upon its number of nucleons. This infers that a single potential can describe the interaction of particles with different energies with different nuclei.

  11. Advances in Zero-Field Nuclear Magnetic Resonance Spectroscopy

    OpenAIRE

    Theis, Thomas

    2012-01-01

    In the course of the last century, Nuclear magnetic resonance (NMR) has become a powerful and ubiquitous analytical tool for the determination of molecular identity, structure, and function. Traditionally, the great analytical power of NMR comes at the cost of mobility and large expenses for cryogenic cooling. This thesis presents how zero-field NMR detected with an atomic magnetometer is emerging as a new, potentially portable and cost-effective modality of NMR with the ability of providing ...

  12. Basic Research Needs for Advanced Nuclear Systems. Report of the Basic Energy Sciences Workshop on Basic Research Needs for Advanced Nuclear Energy Systems, July 31-August 3, 2006

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, J.; Diaz de la Rubia, T.; Gibala, R.; Zinkle, S.; Miller, J.R.; Pimblott, S.; Burns, C.; Raymond, K.; Grimes, R.; Pasamehmetoglu, K.; Clark, S.; Ewing, R.; Wagner, A.; Yip, S.; Buchanan, M.; Crabtree, G.; Hemminger, J.; Poate, J.; Miller, J.C.; Edelstein, N.; Fitzsimmons, T.; Gruzalski, G.; Michaels, G.; Morss, L.; Peters, M.; Talamini, K.

    2006-10-01

    The global utilization of nuclear energy has come a long way from its humble beginnings in the first sustained nuclear reaction at the University of Chicago in 1942. Today, there are over 440 nuclear reactors in 31 countries producing approximately 16% of the electrical energy used worldwide. In the United States, 104 nuclear reactors currently provide 19% of electrical energy used nationally. The International Atomic Energy Agency projects significant growth in the utilization of nuclear power over the next several decades due to increasing demand for energy and environmental concerns related to emissions from fossil plants. There are 28 new nuclear plants currently under construction including 10 in China, 8 in India, and 4 in Russia. In the United States, there have been notifications to the Nuclear Regulatory Commission of intentions to apply for combined construction and operating licenses for 27 new units over the next decade. The projected growth in nuclear power has focused increasing attention on issues related to the permanent disposal of nuclear waste, the proliferation of nuclear weapons technologies and materials, and the sustainability of a once-through nuclear fuel cycle. In addition, the effective utilization of nuclear power will require continued improvements in nuclear technology, particularly related to safety and efficiency. In all of these areas, the performance of materials and chemical processes under extreme conditions is a limiting factor. The related basic research challenges represent some of the most demanding tests of our fundamental understanding of materials science and chemistry, and they provide significant opportunities for advancing basic science with broad impacts for nuclear reactor materials, fuels, waste forms, and separations techniques. Of particular importance is the role that new nanoscale characterization and computational tools can play in addressing these challenges. These tools, which include DOE synchrotron X

  13. Improving human reliability through better nuclear power plant system design: Program for advanced nuclear power studies. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Golay, M.W.

    1993-10-10

    The project on ``Development of a Theory of the Dependence of Human Reliability upon System Designs as a Means of Improving Nuclear Power Plant Performance`` was been undertaken in order to address the problem of human error in advanced nuclear power plant designs. Lack of a mature theory has retarded progress in reducing likely frequencies of human errors. Work being pursued in this project is to perform a set of experiments involving human subjects who are required to operate, diagnose and respond to changes in computer-simulated systems, relevant to those encountered in nuclear power plants, which are made to differ in complexity in a systematic manner. The computer program used to present the problems to be solved also records the response of the operator as it unfolds.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

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

  15. Advanced maintenance, inspection & repair technology for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hinton, B.M.

    1994-12-31

    Maintenance, inspection, and repair technology for nuclear power plants is outlined. The following topics are discussed: technology for reactor systems, reactor refueling bridge, fuel inspection system, fuel shuffling software, fuel reconstitution, CEA/RCCA/CRA inspection, vessel inspection capabilities, CRDM inspection and repair, reactor internals inspection and repair, stud tensioning system, stud/nut cleaning system, EDM machining technology, MI Cable systems, core exit T/C nozzle assemblies, technology for steam generators, genesis manipulator systems, ECT, UT penetrant inspections, steam generator repair and cleaning systems, technology for balance of plant, heat exchangers, piping and weld inspections, and turbogenerators.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-01

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

  17. Radioprotection of the environment. Recent advances in nuclear ecotoxicology research

    Energy Technology Data Exchange (ETDEWEB)

    Garnier-Laplace, J.; Gilbin, R. [Institut de Radioprotection et de Surete Nucleaire, Service d' Etudes du Comportement des Radionucleides dans les Ecosystemes (IRSN/DEI/SECRE), Saint-Paul-Lez-Durance (France); Alonzo, F.; Beaugelin, K.; Hinton, T.G.

    2010-07-01

    Funding for radioecology in Europe escalated after the Chernobyl accident, and remained elevated for some 15 year. The enhanced funding permitted Europeans to explore new directions in radioecology, particularly in the area of effects to non-human biota. Herein we highlight several recent advances: (1) an attempt to merge environmental risk analysis methods for radioactive contamination with those for other pollutants using species sensitivity distribution; (2) attempts to extrapolate damage observed in individual organisms to potential effects in their populations; and (3) the use of advanced models to estimate and explain changes in an individual's energy allocation as a result of exposure to low doses of radionuclides. We conclude by presenting the radioecology Alliance: an international network whose goal is to integrate resources in order to efficiently fill knowledge gaps in radioecology and improve risk assessments tools that support both human and environmental radioprotection. (orig.)

  18. Technology Roadmap Instrumentation, Control, and Human-Machine Interface to Support DOE Advanced Nuclear Energy Programs

    Energy Technology Data Exchange (ETDEWEB)

    Donald D Dudenhoeffer; Burce P Hallbert

    2007-03-01

    Instrumentation, Controls, and Human-Machine Interface (ICHMI) technologies are essential to ensuring delivery and effective operation of optimized advanced Generation IV (Gen IV) nuclear energy systems. In 1996, the Watts Bar I nuclear power plant in Tennessee was the last U.S. nuclear power plant to go on line. It was, in fact, built based on pre-1990 technology. Since this last U.S. nuclear power plant was designed, there have been major advances in the field of ICHMI systems. Computer technology employed in other industries has advanced dramatically, and computing systems are now replaced every few years as they become functionally obsolete. Functional obsolescence occurs when newer, more functional technology replaces or supersedes an existing technology, even though an existing technology may well be in working order.Although ICHMI architectures are comprised of much of the same technology, they have not been updated nearly as often in the nuclear power industry. For example, some newer Personal Digital Assistants (PDAs) or handheld computers may, in fact, have more functionality than the 1996 computer control system at the Watts Bar I plant. This illustrates the need to transition and upgrade current nuclear power plant ICHMI technologies.

  19. Measurements of neutron cross sections for advanced nuclear energy systems at n_TOF (CERN

    Directory of Open Access Journals (Sweden)

    Barbagallo M.

    2014-03-01

    Full Text Available The n_TOF facility operates at CERN with the aim of addressing the request of high accuracy nuclear data for advanced nuclear energy systems as well as for nuclear astrophysics. Thanks to the features of the neutron beam, important results have been obtained on neutron induced fission and capture cross sections of U, Pu and minor actinides. Recently the construction of another beam line has started; the new line will be complementary to the first one, allowing to further extend the experimental program foreseen for next measurement campaigns.

  20. Recent advances in echocardiography for nuclear medicine physician

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Geu Ru; Shin, Dong Gu [Yeungnam University College of Medicine, Daegu (Korea, Republic of)

    2005-10-15

    Echocardiography is one of the most frequently used techniques for diagnosing cardiovascular diseases. Over the last twenty years, technological advances have enabled the application of high-quality imaging. Important recent developments have occurred in echocardiography that are already being used clinically. Equipment and hardware is now available to produce real time three-dimensional and contrast enhanced imaging. Tissue Doppler and stress echocardiography have provided potential benefit to analyze hemodynamic information of heart. This review discusses each of these new developments and their potential impact on the practice of echocardiography and cardiology in general.

  1. Nanostructured Fe-Cr Alloys for Advanced Nuclear Energy Applications

    Energy Technology Data Exchange (ETDEWEB)

    Scattergood, Ronald O. [North Carolina State Univ., Raleigh, NC (United States)

    2016-04-26

    We have completed research on the grain-size stabilization of model nanostructured Fe14Cr base alloys at high temperatures by the addition of non-equilibrium solutes. Fe14Cr base alloys are representative for nuclear reactor applications. The neutron flux in a nuclear reactor will generate He atoms that coalesce to form He bubbles. These can lead to premature failure of the reactor components, limiting their lifetime and increasing the cost and capacity for power generation. In order to mitigate such failures, Fe14Cr base alloys have been processed to contain very small nano-size oxide particles (less than 10 nm in size) that trap He atoms and reduce bubble formation. Theoretical and experimental results indicate that the grain boundaries can also be very effective traps for He atoms and bubble formation. An optimum grain size will be less than 100 nm, ie., nanocrystalline alloys must be used. Powder metallurgy methods based on high-energy ball milling can produce Fe-Cr base nanocrystalline alloys that are suitable for nuclear energy applications. The problem with nanocrystalline alloys is that excess grain-boundary energy will cause grains to grow at higher temperatures and their propensity for He trapping will be lost. The nano-size oxide particles in current generation nuclear alloys provide some grain size stabilization by reducing grain-boundary mobility (Zener pinning – a kinetic effect). However the current mitigation strategy minimizing bubble formation is based primarily on He trapping by nano-size oxide particles. An alternate approach to nanoscale grain size stabilization has been proposed. This is based on the addition of small amounts of atoms that are large compared to the base alloy. At higher temperatures these will diffuse to the grain boundaries and will produce an equilibrium state for the grain size at higher temperatures (thermodynamic stabilization – an equilibrium effect). This would be preferred compared to a kinetic effect, which is not

  2. Recent advancements in cloning by somatic cell nuclear transfer.

    Science.gov (United States)

    Ogura, Atsuo; Inoue, Kimiko; Wakayama, Teruhiko

    2013-01-05

    Somatic cell nuclear transfer (SCNT) cloning is the sole reproductive engineering technology that endows the somatic cell genome with totipotency. Since the first report on the birth of a cloned sheep from adult somatic cells in 1997, many technical improvements in SCNT have been made by using different epigenetic approaches, including enhancement of the levels of histone acetylation in the chromatin of the reconstructed embryos. Although it will take a considerable time before we fully understand the nature of genomic programming and totipotency, we may expect that somatic cell cloning technology will soon become broadly applicable to practical purposes, including medicine, pharmaceutical manufacturing and agriculture. Here we review recent progress in somatic cell cloning, with a special emphasis on epigenetic studies using the laboratory mouse as a model.

  3. Gas cooling in semi-analytic models and smoothed particle hydrodynamics simulations: are results consistent?

    Science.gov (United States)

    Saro, A.; De Lucia, G.; Borgani, S.; Dolag, K.

    2010-08-01

    We present a detailed comparison between the galaxy populations within a massive cluster, as predicted by hydrodynamical smoothed particle hydrodynamics (SPH) simulations and by a semi-analytic model (SAM) of galaxy formation. Both models include gas cooling and a simple prescription of star formation, which consists in transforming instantaneously any cold gas available into stars, while neglecting any source of energy feedback. This simplified comparison is thus not meant to be compared with observational data, but is aimed at understanding the level of agreement, at the stripped-down level considered, between two techniques that are widely used to model galaxy formation in a cosmological framework and which present complementary advantages and disadvantages. We find that, in general, galaxy populations from SAMs and SPH have similar statistical properties, in agreement with previous studies. However, when comparing galaxies on an object-by-object basis, we find a number of interesting differences: (i) the star formation histories of the brightest cluster galaxies (BCGs) from SAM and SPH models differ significantly, with the SPH BCG exhibiting a lower level of star formation activity at low redshift, and a more intense and shorter initial burst of star formation with respect to its SAM counterpart; (ii) while all stars associated with the BCG were formed in its progenitors in the SAM used here, this holds true only for half of the final BCG stellar mass in the SPH simulation, the remaining half being contributed by tidal stripping of stars from the diffuse stellar component associated with galaxies accreted on the cluster halo; (iii) SPH satellites can lose up to 90 per cent of their stellar mass at the time of accretion, due to tidal stripping, a process not included in the SAM used in this paper; (iv) in the SPH simulation, significant cooling occurs on the most massive satellite galaxies and this lasts for up to 1 Gyr after accretion. This physical process is

  4. Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Frisani, Angelo; Hassan, Yassin A; Ugaz, Victor M

    2010-11-02

    The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the

  5. Cladding and Structural Materials for Advanced Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Was, G S; Allen, T R; Ila, D; C,; Levi,; Morgan, D; Motta, A; Wang, L; Wirth, B

    2011-06-30

    The goal of this consortium is to address key materials issues in the most promising advanced reactor concepts that have yet to be resolved or that are beyond the existing experience base of dose or burnup. The research program consists of three major thrusts: 1) high-dose radiation stability of advanced fast reactor fuel cladding alloys, 2) irradiation creep at high temperature, and 3) innovative cladding concepts embodying functionally-graded barrier materials. This NERI-Consortium final report represents the collective efforts of a large number of individuals over a period of three and a half years and included 9 PIs, 4 scientists, 3 post-docs and 12 students from the seven participating institutions and 8 partners from 5 national laboratories and 3 industrial institutions (see table). University participants met semi-annually and participants and partners met annually for meetings lasting 2-3 days and designed to disseminate and discuss results, update partners, address outstanding issues and maintain focus and direction toward achieving the objectives of the program. The participants felt that this was a highly successful program to address broader issues that can only be done by the assembly of a range of talent and capabilities at a more substantial funding level than the traditional NERI or NEUP grant. As evidence of the success, this group, collectively, has published 20 articles in archival journals and made 57 presentations at international conferences on the results of this consortium.

  6. Nuclear systems

    CERN Document Server

    Todreas, Neil E

    2011-01-01

    Principal Characteristics of Power ReactorsIntroductionPower CyclesPrimary Coolant SystemsReactor CoresFuel AssembliesAdvanced Water- and Gas-Cooled Reactors (Generation III And III+)Advanced Thermal and Fast Neutron Spectrum Reactors (Generation IV)ReferencesProblemsThermal Design Principles and ApplicationIntroductionOverall Plant Characteristics Influenced by Thermal Hydraulic ConsiderationsEnergy Production and Transfer ParametersThermal Design LimitsThermal Design MarginFigures of Merit for Core Thermal PerformanceThe Inverted Fuel ArrayThe Equivalent Annulus ApproximationReferencesProble

  7. Nuclear β-catenin expression as a prognostic factor in advanced colorectal carcinoma

    Institute of Scientific and Technical Information of China (English)

    Adam Elzagheid; Abdelbaset Buhmeida; Eija Korkeila; Yrj(o) Collan; Karl Syrj(a)nen; Seppo Pyrh(o)nen

    2008-01-01

    AIM: To investigate the changing pattern of β-catenin expression and its prognostic value in advanced colorectal cancer (CRC).METHODS.Archival tumor samples were analyzed for β-catenin using immunohistochemisry (IHC) in 95 patients with advanced CRC.RESULTS: Membranous β-catenin expression was found in the normal colorectal epithelium.Almost 100% of CRCcases showed membranous and cytoplasmic expression,and 55 (58%) cases showed nuclear expression.In univariate (Kaplan-Meier)survival analysis,only the nuclear index (NI) was a significant predictor of disease free survival (DFS) (P=0.023; n = 35),with a NI above the median associated with longer DFS (34.2 mo) than those with a NI below the median (15.5 mo) (P = 0.045,ANOVA).The other indices were not significant predictors of DFS,and none of the three tested indices (for membranous,cytoplasmic,or nuclear expression) predicted diseasespecific survival (DSS).However,when dichotomized as positive or negative nuclear expression,the former was a significant predictor of more favorable DFS (P =0.041) and DSS (P = 0.046).CONCLUSION: Nuclear β-catenin expression provides additional information in predicting patient outcome in advanced CRC.

  8. UNEDF: Advanced Scientific Computing Transforms the Low-Energy Nuclear Many-Body Problem

    CERN Document Server

    Stoitsov, M; Nazarewicz, W; Bulgac, A; Hagen, G; Kortelainen, M; Pei, J C; Roche, K J; Schunck, N; Thompson, I; Vary, J P; Wild, S M

    2011-01-01

    The UNEDF SciDAC collaboration of nuclear theorists, applied mathematicians, and computer scientists is developing a comprehensive description of nuclei and their reactions that delivers maximum predictive power with quantified uncertainties. This paper illustrates significant milestones accomplished by UNEDF through integration of the theoretical approaches, advanced numerical algorithms, and leadership class computational resources.

  9. AP1000, a nuclear central of advanced design; AP1000, una central nuclear de diseno avanzado

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, N.; Viais J, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: nhm@nuclear.inin.mx

    2005-07-01

    The AP1000 is a design of a nuclear reactor of pressurized water (PWR) of 1000 M We with characteristic of safety in a passive way; besides presenting simplifications in the systems of the plant, the construction, the maintenance and the safety, the AP1000 is a design that uses technology endorsed by those but of 30 years of operational experience of the PWR reactors. The program AP1000 of Westinghouse is focused to the implementation of the plant to provide improvements in the economy of the same one and it is a design that is derived directly of the AP600 designs. On September 13, 2004 the US-NRC (for their initials in United States- Nuclear Regulatory Commission) approved the final design of the AP1000, now Westinghouse and the US-NRC are working on the whole in a complete program for the certification. (Author)

  10. Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hui Zhang; Raman P. Singh

    2008-11-30

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor componets is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.

  11. Systematic Discrimination of Advanced Hydrogen Production Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Charles V. Park; Michael W. Patterson

    2010-07-01

    The U.S. Department of Energy, in concert with industry, is developing a high-temperature gas-cooled reactor at the Idaho National Laboratory (INL) to demonstrate high temperature heat applications to produce hydrogen and electricity or to support other industrial applications. A key part of this program is the production of hydrogen from water that would significantly reduce carbon emissions compared to current production using natural gas. In 2009 the INL led the methodical evaluation of promising advanced hydrogen production technologies in order to focus future resources on the most viable processes. This paper describes how the evaluation process was systematically planned and executed. As a result, High-Temperature Steam Electrolysis was selected as the most viable near-term technology to deploy as a part of the Next Generation Nuclear Plant Project.

  12. Strategic Plan for Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

    Energy Technology Data Exchange (ETDEWEB)

    Kimberlyn C. Mousseau

    2011-10-01

    The Nuclear Energy Computational Fluid Dynamics Advanced Modeling and Simulation (NE-CAMS) system is being developed at the Idaho National Laboratory (INL) in collaboration with Bettis Laboratory, Sandia National Laboratory (SNL), Argonne National Laboratory (ANL), Utah State University (USU), and other interested parties with the objective of developing and implementing a comprehensive and readily accessible data and information management system for computational fluid dynamics (CFD) verification and validation (V&V) in support of nuclear energy systems design and safety analysis. The two key objectives of the NE-CAMS effort are to identify, collect, assess, store and maintain high resolution and high quality experimental data and related expert knowledge (metadata) for use in CFD V&V assessments specific to the nuclear energy field and to establish a working relationship with the U.S. Nuclear Regulatory Commission (NRC) to develop a CFD V&V database, including benchmark cases, that addresses and supports the associated NRC regulations and policies on the use of CFD analysis. In particular, the NE-CAMS system will support the Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program, which aims to develop and deploy advanced modeling and simulation methods and computational tools for reliable numerical simulation of nuclear reactor systems for design and safety analysis. Primary NE-CAMS Elements There are four primary elements of the NE-CAMS knowledge base designed to support computer modeling and simulation in the nuclear energy arena as listed below. Element 1. The database will contain experimental data that can be used for CFD validation that is relevant to nuclear reactor and plant processes, particularly those important to the nuclear industry and the NRC. Element 2. Qualification standards for data evaluation and classification will be incorporated and applied such that validation data sets will result in well

  13. UKAEA's role in the future of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.

    1988-01-01

    The article on how the UKAEA sees its role in the future was first presented at the conference on 'The future of nuclear power after Sizewell B', held in London, 1987. A description is given of the context within which the UKAEA works. To meet its objectives and the needs of its customers, the authority plans to extend its studies along three main lines of development. The first is the development of future nuclear energy systems, including the fast reactor and fusion. The second is the support of the current United Kingdom nuclear technology, which includes work on:-magnox reactors, advanced gas cooled reactors, fuel cycle research, waste disposal and decommissioning. The third is on underpinning research and cross-fertilisation.

  14. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    Science.gov (United States)

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

  15. Assessment of fiber optic sensors and other advanced sensing technologies for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hashemian, H.M. [Analysis and Measurement Services Corporation, Knoxville, TN (United States)

    1996-03-01

    As a result of problems such as calibration drift in nuclear plant pressure sensors and the recent oil loss syndrome in some models of Rosemount pressure transmitters, the nuclear industry has become interested in fiber optic pressure sensors. Fiber optic sensing technologies have been considered for the development of advanced instrumentation and control (I&C) systems for the next generation of reactors and in older plants which are retrofitted with new I&C systems. This paper presents the results of a six-month Phase I study to establish the state-of-the-art in fiber optic pressure sensing. This study involved a literature review, contact with experts in the field, an industrial survey, a site visit to a fiber optic sensor manufacturer, and laboratory testing of a fiber optic pressure sensor. The laboratory work involved both static and dynamic performance tests. This initial Phase I study has recently been granted a two-year extension by the U.S. Nuclear Regulatory Commission (NRC). The next phase will evaluate fiber optic pressure sensors in specific nuclear plant applications in addition to other advanced methods for monitoring critical nuclear plant equipment.

  16. A Study on intensifying efficiency for international collaborative development of Advanced Nuclear Energy Technology

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Moon Hee; Kim, H. R.; Kim, H. J.; Chang, J. H.; Hahn, D. H.; Bae, Y. Y.; Kim, W. W.; Jeong, I.; Lee, D. S.; Lee, J. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-06-15

    Generation IV International Forum(GIF), where 13 countries including Korea collaborate to develop future nuclear energy systems, put into force 'Generation IV International Forum Project Arrangement' in 2007 for the international research and development of Gen IV Systems, following the entry into force of Framework Agreement in 2005. The International Nuclear Research Initiative(I-NERI) between Korea and United States and the International Project on Innovative Nuclear Energy Systems and Fuel Cycles(INPRO) of IAEA are continued in this year, produced lots of visible outcomes. These international activities have a common goal of the collaborative development of advanced nuclear system technologies but differ in the main focusing areas and aspects, so Korea needs to establish the integrated strategy based on the distinguished and complementary approach for the participation of each international programs, as examples the GIF for the advanced system technology development, INPRO for the set-up of institution and infra-structure, and I-NERI for the access of the core technologies and acquisition of the transparency of nuclear R and D.

  17. Advanced Accelerator Applications University Participation Program

    Energy Technology Data Exchange (ETDEWEB)

    Y. Chen; A. Hechanova

    2007-07-25

    Our research tasks span the range of technology areas for transmutation, gas-cooled reactor technology, and high temperature heat exchangers, including separation of actinides from spent nuclear fuel, methods of fuel fabrication, reactor-accelerator coupled experiments, corrosion of materials exposed to lead-bismuth eutectic, and special nuclear materials protection and accountability.

  18. Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, Hiroyuki, E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, Hideaki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Katayama, Kazunari [Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-koen, Kasuga 8168580 (Japan); Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan)

    2015-10-15

    Highlights: • Tritium production by a high-temperature gas-cooled reactor was studied. • The loading method considering tritium outflow suppression was estimated. • A reactor with 600 MWt produced 400–600 g of tritium for 180 days. • A possibility that tritium outflow can be sufficiently suppressed was shown. - Abstract: Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al{sub 2}O{sub 3} cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300 MW electrical nominal capacity (GTHTR300) with 600 MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600 g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-30

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

  20. Recent Advances in Computational Methods for Nuclear Magnetic Resonance Data Processing

    KAUST Repository

    Gao, Xin

    2013-01-11

    Although three-dimensional protein structure determination using nuclear magnetic resonance (NMR) spectroscopy is a computationally costly and tedious process that would benefit from advanced computational techniques, it has not garnered much research attention from specialists in bioinformatics and computational biology. In this paper, we review recent advances in computational methods for NMR protein structure determination. We summarize the advantages of and bottlenecks in the existing methods and outline some open problems in the field. We also discuss current trends in NMR technology development and suggest directions for research on future computational methods for NMR.

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

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

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

  2. Recent advances of annular centrifugal extractor for hot test of nuclear waste partitioning process

    Institute of Scientific and Technical Information of China (English)

    HeXiang-Ming; YanYu-Shun; 等

    1998-01-01

    Advances are being made in the design of the annular centrifugal extractor fornuclear fuel reprocessing extraction process studies.The extractors have been built and tested.Twelve stages of this extractor and 50 stages are used toimplement the TRPO process for the cleanup ofcommercial and defense nuclear waste liquids,respectively.Following advances are available:(1) simple way of assembly and disassembly between rotor part and housing part of extractor,ease of manipulator operation;(2)automatic sampling from housing of extractor in hot cell;(3) compact multi-stage housing system;(4) easy interstage link;(5) computer data acquisition and monitoring system of speed.

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

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

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

  4. The development of advanced robotics for the nuclear industry -The development of advanced robotic technology-

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Min; Lee, Yong Bum; Park, Soon Yong; Cho, Jae Wan; Lee, Nam Hoh; Kim, Woong Kee; Moon, Byung Soo; Kim, Seung Hoh; Kim, Chang Heui; Kim, Byung Soo; Hwang, Suk Yong; Lee, Yung Kwang; Moon, Je Sun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    Main activity in this year is to develop both remote handling system and telepresence techniques, which can be used for people involved in extremely hazardous working area to alleviate their burden. In the robot vision technology part, KAERI-PSM system, stereo imaging camera module, stereo BOOM/MOLLY unit, and stereo HMD unit are developed. Also, autostereo TV system which falls under the category of next generation stereo imaging technology has been studied. The performance of KAERI-PSM system for remote handling task is evaluated and compared with other stereo imaging systems as well as general TV imaging system. The result shows that KAERI-PSM system is superior to the other stereo imaging systems about remote operation speedup and accuracy. The automatic recognition algorithm of instrument panel is studied and passive visual target tracking system is developed. The 5 DOF camera serving unit has been designed and fabricated. It is designed to function like human`s eye. In the sensing and intelligent control research part, thermal image database system for thermal image analysis is developed and remote temperature monitoring technique using fiber optics is investigated. And also, two dimensional radioactivity sensor head for radiation profile monitoring system is designed. In the part of intelligent robotics, mobile robot is fabricated and its autonomous navigation using fuzzy control logic is studied. These remote handling and telepresence techniques developed in this project can be applied to nozzle-dam installation/removal robot system, reactor inspection unit, underwater nuclear pellet inspection and pipe abnormality inspection. And these developed remote handling and telepresence techniques will be applied in general industry, medical science, and military as well as nuclear facilities. 203 figs, 12 tabs, 72 refs. (Author).

  5. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre, Institut für Transurane, Hermann-von-Helmholtz-Platz 1, 76344, Eggenstein-Leopoldshafen (Germany); Fourrez, S. [THERMOCOAX SAS, Rue du Pré Neuf, 61100 Saint Georges des Groseillers (France); Fütterer, M.A.; Lapetite, J.M. [European Commission, Joint Research Centre, Institut für Transurane, Hermann-von-Helmholtz-Platz 1, 76344, Eggenstein-Leopoldshafen (Germany); Sadli, M.; Morice, R.; Failleau, G. [Laboratoire commun de métrologie LNE-Cnam, 61 rue du Landy, F-92310 Saint-Denis (France)

    2014-05-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innovative sheathed thermocouples which would consecutively be tested under irradiation. Two such screening tests have been performed in high temperature environment (i.e. temperature in the range 1100–1150 °C) with purposely contaminated helium atmosphere (mainly CH{sub 4}, CO, CO{sub 2}, O{sub 2} impurities) representative for high temperature reactor carburizing atmospheres. The first set of thermocouples embedded in graphite (mainly conventional N type thermocouples and thermocouples with innovative sheaths) was tested in a dedicated furnace at THERMOCOAX lab with helium flushing. The second out-of-pile test at JRC with a partly different set of thermocouples replicated the original test for comparison. Performance indicators such as thermal drift, insulation resistance and loop resistance were monitored. Through these long-term screening tests the effect of several parameters were investigated: niobium sleeves, bending, diameter, sheath composition as well as the chemical environment. SEM examinations were performed to analyze local damage (bending zone, sheath). The present paper describes the two tests, sums up data collected during these tests in terms of thermocouple behavior and describes further instrumentation testing work with fixed point mini cells for qualification under irradiation.

  6. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Laurie, M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, F-61438 Flers Cedex (France); Fuetterer, M. A.; Lapetite, J. M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

    2011-07-01

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remedy this problem, out-of-pile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of thermocouples embedded in graphite (mainly conventional Type N thermocouples and thermocouples with innovative sheaths) in a dedicated furnace with helium flushing. Performance indicators such as thermal drift, insulation and loop resistance were monitored and compared to those from conventional Type N thermocouples. Several parameters were investigated: niobium sleeves, bending, thickness, sheath composition, temperature as well as the chemical environment. After the tests, Scanning Electron Microscopy (SEM) examinations were performed to analyze possible local damage in wires and in the sheath. The present paper describes the two experiments, summarizes results and outlines further work, in particular to further analyze the findings and to select suitable thermocouples for qualification under irradiation. (authors)

  7. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stéphane; FUETTERER Michael; LAPETITE Jean-Marc; SADLI M.; MORICE Ronan; FAILLEAU G

    2013-01-01

    During irradiation tests at high temperature failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. As instrumentation, in particular thermocouples are considered safety-relevant both for irradiation tests and for commercial reactors, JRC and THERMOCOAX joined forces to solve this issue by performing out-of-pile tests with thermocouples mimicking the environment encountered by high temperature reactor (HTR) in-core instrumentation. The objective was to screen innova...

  8. Long Term Out-of-pile Thermocouple Tests in Conditions Representative for Nuclear Gas-cooled High Temperature Reactors

    OpenAIRE

    LAURIE Mathias; FOURREZ Stephane; FUETTERER Michael; LAPETITE Jean-Marc

    2011-01-01

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remediate this problem, out-ofpile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of...

  9. A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators

    Energy Technology Data Exchange (ETDEWEB)

    White, L.S.

    1990-07-01

    This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.

  10. Advanced international training course on state systems of accounting for and control of nuclear materials

    Energy Technology Data Exchange (ETDEWEB)

    1981-10-01

    This report incorporates all lectures and presentations at the Advanced International Training Course on State Systems of Accounting for and Control of Nuclear Material held April 27 through May 12, 1981 at Santa Fe and Los Alamos, New Mexico, and Richland, Washington, USA. Authorized by the US Nuclear Non-Proliferation Act and sponsored by the US Department of Energy in cooperation with the International Atomic Energy Agency, the course was developed to provide practical training in the design, implementation, and operation of a state system of nuclear materials accountability and control that satisfies both national and international safeguards. Major emphasis for the 1981 course was placed on safeguards methods used at bulk-handling facilities, particularly low-enriched uranium conversion and fuel fabrication plants. The course was conducted by the University of California's Los Alamos National Laboratory, the Battelle Pacific Northwest Laboratory, and Exxon Nuclear Company, Inc. Tours and demonstrations were arranged at both the Los Alamos National Laboratory, Los Alamos, New Mexico, and the Exxon Nuclear fuel fabrication plant, Richland, Washington.

  11. NATO Advanced Research Workshop “Nuclear Power and Energy Security”

    CERN Document Server

    Apikyan, Samuel A; Nuclear Power and Energy Security

    2010-01-01

    World energy consumption has grown dramatically over the past few decades. This growth in energy demand will be driven by large increases in both economic growth and world population coupled with rising living standards in rapidly growing countries. The last years, we routinely hear about a "renaissance" of nuclear energy. The recognition that nuclear power is vital to global energy security in the 21st century has been growing for some time. "The more we look to the future, the more we can expect countries to be considering the potential benefits that expanding nuclear power has to offer for the global environment and for economic growth," IAEA Director General Mohamed ElBaradei said in advance of a gathering of 500 nuclear power experts assembled in Moscow for the "International Conference on Fifty Years of Nuclear Power - the Next Fifty Years". But such a renaissance is not a single-valued and sure thing. Legitimate four unresolved questions remain about high relative costs; perceived adverse safety, envir...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-29

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

  13. Providing Policy Implication Based on the R and D Portfolio Analysis in Advanced Countries in the Nuclear Technology

    Energy Technology Data Exchange (ETDEWEB)

    Moon, K. H.; Lee, M. K.; Won, B. C.; Kim, S. S.; Lee, J. H.; Yun, S. W.; Jeong, I. K.; Lee, Y. C.; Lee, Y. J.; Kim, Y. S.

    2013-08-15

    This study is to provide the investment direction of nuclear R and D, which is the most efficient and reasonable integrating the various aspects comprehensively. This study includes four parts. In the first part, we extracted Mega-trend and driving forces of nuclear R and D field by using various reports published by National Intelligence Council of US, UN, etc. Also, in this part we established the linkage between megatrend factors focussing on nuclear and the five aspects including society, technology, ecology, economics and politics. In the second part, we analyzed the nuclear R and D investment directions of major advanced countries including US, Japan, EU and China for comparing the investment portfolio in the specific research area. In the third part, domestic investment of nuclear R and D was reviewed by analyzing the investment trend of nuclear R and D in the past, with their connection to nuclear policy, and to the levels and capacities of national technologies of nuclear. In the final part, the desirable directions of nuclear R and D investment were suggested comprehensively taking into consideration various aspects including the Mega-trend associated with nuclear, nuclear R and D directions of major advanced countries, and the level and capacities of the domestic nuclear technologies.

  14. Alternative nuclear technologies

    Science.gov (United States)

    Schubert, E.

    1981-10-01

    The lead times required to develop a select group of nuclear fission reactor types and fuel cycles to the point of readiness for full commercialization are compared. Along with lead times, fuel material requirements and comparative costs of producing electric power were estimated. A conservative approach and consistent criteria for all systems were used in estimates of the steps required and the times involved in developing each technology. The impact of the inevitable exhaustion of the low- or reasonable-cost uranium reserves in the United States on the desirability of completing the breeder reactor program, with its favorable long-term result on fission fuel supplies, is discussed. The long times projected to bring the most advanced alternative converter reactor technologies the heavy water reactor and the high-temperature gas-cooled reactor into commercial deployment when compared to the time projected to bring the breeder reactor into equivalent status suggest that the country's best choice is to develop the breeder. The perceived diversion-proliferation problems with the uranium plutonium fuel cycle have workable solutions that can be developed which will enable the use of those materials at substantially reduced levels of diversion risk.

  15. Analysis on thermophoretic deposit of fine particle on water wall of 10 MW high temperature gas-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    ZHOU Tao; YANG Rui-Chang; JIA Dou-Nan

    2005-01-01

    The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calculated using annular vertical closed cavity model. Fine particles can deposit on the water wall due to the thermophore sis effect. This deposit can affect heat transfer. The thermophoretic deposit efficiency is calculated by using Batch and Shen's formula fitted for both laminar flow and turbulent flow. The calculated results indicate that natural convection is turbulent in the closed cavity. The transient thermophoretic deposit efficiency rises with the increase of the pressure shell's temperature. Its maximum value is 14%.

  16. Analysis of two-phase flow instability in helical tube steam generator in high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Yu; Lv, Xuefeng; Wang, Shengfei; Niu, Fenglei; Tian, Li [North China Electric Power Univ., Beijing (Switzerland)

    2012-03-15

    The steam generator composed of multi-helical tubes is used in high temperature gas cooled reactors and two-phase flow instability should be avoided in design. And density-wave oscillation which is mainly due to flow, density and the relationship between the pressure drop delays and feedback effects is one of the two-phase flow instability phenomena easily to occur. Here drift-flux model is used to simulate the performance of the fluid in the secondary side and frequency domain and time domain methods are used to evaluate whether the density-wave oscillation will happen or not. Several operating conditions with nominal power from 15% to 30% are calculated in this paper. The results of the two methods are in accordance, flow instability will occur when power is less than 20% nominal power, which is also according with the result of the experiments well.

  17. Studies on disintegrating spherical fuel elements of high temperature gas-cooled reactor by a electrochemical method

    Science.gov (United States)

    Tian, Lifang; Wen, Mingfen; Chen, Jing

    2013-01-01

    Spherical fuel elements of a high temperature gas-cooled reactor were disintegrated through a electrochemical method with NaNO3 as electrolyte. The X-ray diffraction spectra and total carbon contents of the graphite fragments were determined, and the results agreed with those from simulated fuel elements. After conducting the characterization analysis and the leaching experiment of coated fuel particles, the uranium concentrations of leaching solutions and spent electrolyte were found to be at background levels. The results demonstrate the effectiveness of the improved electrochemical method with NaNO3 as electrolyte in disintegrating the unirradiated fuel elements without any damage to the coated fuel particles. Moreover, the method avoided unexpected radioactivity contamination to the graphite matrix and spent electrolyte.

  18. CFD Analysis of Turbulent Flow Phenomena in the Lower Plenum of a Prismatic Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    T. Gallaway; S.P. Antal; M.Z. Podowski; D.P. Guillen

    2007-09-01

    This paper is concerned with the implementation of a computational model of turbulent flow in a section of the lower plenum of Very High Temperature Reactor (VHTR). The proposed model has been encoded in a state-of-the-art CFD code, NPHASE. The results of NPHASE predictions have been compared against the experimental data collected using a scaled model of a sub-region in the lower plenum of a modular prismatic gas-cooled reactor. It has been shown that the NPHASE-based model is capable of predicting a three-dimensional velocity field in a complex geometrical configuration of VHTR lower plenum. The current and future validations of computational predictions are necessary for design and analysis of new reactor concepts, as well as for safety analysis and licensing calculations.

  19. A cryogenic buffer gas cooled beam of BaH for molecular laser cooling and ultracold fragmentation

    Science.gov (United States)

    Iwata, Geoffrey; Tarallo, Marco G.; Soerensen, Fabian; Zelevinsky, Tanya

    2015-05-01

    Laser cooled and trapped molecules promise many possibilities to explore a variety of fields such as many-body physics, quantum collisions and dissociation, and precision measurement. We report on an experiment for cooling and trapping barium monohydride (BaH) diatomic molecules. We present a cryogenic buffer gas cooling apparatus for producing a 4 K beam of BaH, and describe the laser cooling schemes necessary to load a molecular magneto-optical trap from that beam. Current progress includes identification of the cooling transitions in the BaH B2 Σ molecules and construction of the molecular beam. The large mass ratio of constituent atoms in BaH makes this system attractive for future studies of ultracold fragmentation, potentially resulting in samples of ultracold hydrogen atoms.

  20. Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2003-02-01

    This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

  1. Strategic Plan for Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

    Energy Technology Data Exchange (ETDEWEB)

    Rich Johnson; Kimberlyn C. Mousseau; Hyung Lee

    2011-09-01

    NE-KAMS knowledge base will assist computational analysts, physics model developers, experimentalists, nuclear reactor designers, and federal regulators by: (1) Establishing accepted standards, requirements and best practices for V&V and UQ of computational models and simulations, (2) Establishing accepted standards and procedures for qualifying and classifying experimental and numerical benchmark data, (3) Providing readily accessible databases for nuclear energy related experimental and numerical benchmark data that can be used in V&V assessments and computational methods development, (4) Providing a searchable knowledge base of information, documents and data on V&V and UQ, and (5) Providing web-enabled applications, tools and utilities for V&V and UQ activities, data assessment and processing, and information and data searches. From its inception, NE-KAMS will directly support nuclear energy research, development and demonstration programs within the U.S. Department of Energy (DOE), including the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Nuclear Energy Advanced Modeling and Simulation (NEAMS), the Light Water Reactor Sustainability (LWRS), the Small Modular Reactors (SMR), and the Next Generation Nuclear Power Plant (NGNP) programs. These programs all involve computational modeling and simulation (M&S) of nuclear reactor systems, components and processes, and it is envisioned that NE-KAMS will help to coordinate and facilitate collaboration and sharing of resources and expertise for V&V and UQ across these programs. In addition, from the outset, NE-KAMS will support the use of computational M&S in the nuclear industry by developing guidelines and recommended practices aimed at quantifying the uncertainty and assessing the applicability of existing analysis models and methods. The NE-KAMS effort will initially focus on supporting the use of computational fluid dynamics (CFD) and thermal hydraulics (T/H) analysis for M&S of nuclear

  2. The development of advanced robotics for the nuclear industry -The development of advanced robotic technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Min; Lee, Yong Bum; Park, Soon Woong; Cho, Jae Wan; Lee, Nam Ho; Kim, Woong Ki; Moon, Byung Soo; Lee, Young Jae; Kim, Chang Hoi; Kim, Seung Ho; Hwang, Seok Yong; Kim, Byung Soo; Moon, Jae Sun; Lee, Young Kwang; Choi, Kap Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    The comparison study of 3 kinds of stereo camera modules done in this final year of 4 year`s longterm project shows that regenerating characteristics of stereo image of stereo camera using horizontally moving lens axis method is superior to the other two modules. Base on this comparison result, stereo camera module using horizontally moving lens method is developed. Also, stereo-Boom unit, high definition polarized stereo monitor(KAERI-PSM II) and 10.4sec. auto-stereogram TV using parallax barrier method are developed. These developed systems can be used for people involved in extremely hazardous working area to give vivid reality image of work environment. In the recognition and tracking section, auto-vergencing technology using focus fixation and cepstral filter, stereo camera calibration, range measurement technology using stereo camera module are developed. And active target tracking technology is developed also. In the sensing and intelligent control research part, active radioactivity image monitoring unit is developed. The spatial resolution of monitoring unit is 10cm at 1m distance, FOV is 60x40 deg [HXV], and radioactivity detection limit is 1mR/hr. Also, radiation-resistant inspection camera for nuclear facilities is designed. In the intelligent control section, fuzzy control algorithm for obstacle detouring navigation of mobile robot is developed. The smoothing techniques by fuzzy set is adapted to raise the pliability of obstacle detouring navigation of mobile robot. In order to raise robustness of developed fuzzy algorithm, fuzzy control algorithm is applied to `Truck Backer Upper` problem and tuned. (Abstract Truncated)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-01

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

  4. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  5. Public perception of the activities of the Nuclear Energy and Advanced Technologies Agency of Cuba

    Energy Technology Data Exchange (ETDEWEB)

    Contreras, Marta; Arencibia, Alois; Alvarado, Jorge; Garcia, Dulce; Rodriguez, Ingrids; Hernandez, Noslen [Centro de Gestion de la Informacion y Desarrollo de la Energia (CUBAENERGIA), La Habana (Cuba); Aguilar, Aurora; Perera, Maricela [Centro de Investigaciones Psicologicas y Sociologicas (CIPS), La Habana, (Cuba); Rodriguez, Ramon [Agencia de Energia Nuclear y Tecnologias de Avanzada (AENTA), La Habana, (Cuba); Alonso, Ivonne [Centro Nacional de Seguridad Nuclear (CNSN), La Habana, (Cuba); Quintana, Natacha [Centro de Aplicaciones Tecnologicas y Desarrollo Nuclear (CEADEN), La Habana, (Cuba); Cardenas, Juan; Ramos, Odalys [Centro de Proteccion e Higiene de las Radiaciones (CPHR), La Habana, (Cuba); Elias, Lidia Lauren [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana, (Cuba)

    2013-07-01

    The work presents the results of a study of perception of the nuclear activities of the Nuclear Energy and Advanced Technologies Agency of Cuba, carried out by means of a study of image. The public object was a wide group of clients, providers, journalists and experts of the governing and regulatory organs which constitute its external public. For the investigation a methodology was prepared with a questionnaire and a semi structured interview, which allowed to obtain complementary information of qualitative character. In general, the perception of the Agency turned out to be positive and the human resources were the best evaluated aspect. Nevertheless, the visibility of the Agency in the public day pupils is considered to be insufficient. The study provided the necessary information so as to design the strategy of communication of the Agency. (author)

  6. Diffusion, Thermal Properties and Chemical Compatibilities of Select MAX Phases with Materials For Advanced Nuclear Systems

    Energy Technology Data Exchange (ETDEWEB)

    Barsoum, Michel [Drexel Univ., Philadelphia, PA (United States); Bentzel, Grady [Drexel Univ., Philadelphia, PA (United States); Tallman, Darin J. [Drexel Univ., Philadelphia, PA (United States); Sindelar, Robert [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Garcia-Diaz, Brenda [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hoffman, Elizabeth [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-04

    The demands of Gen IV nuclear power plants for long service life under neutron irradiation at high temperature are severe. Advanced materials that would withstand high temperatures (up to 1000+ ºC) to high doses in a neutron field would be ideal for reactor internal structures and would add to the long service life and reliability of the reactors. The objective of this work is to investigate the chemical compatibility of select MAX with potential materials that are important for nuclear energy, as well as to measure the thermal transport properties as a function of neutron irradiation. The chemical counterparts chosen for this work are: pyrolytic carbon, SiC, U, Pd, FLiBe, Pb-Bi and Na, the latter 3 in the molten state. The thermal conductivities and heat capacities of non-irradiated MAX phases will be measured.

  7. Army Gas-Cooled Reactor Systems program: alternator final design report

    Energy Technology Data Exchange (ETDEWEB)

    1964-06-01

    The development and testing of a demonstration brushless alternator for the ML-1 mobile nuclear power plant is described. The brushless concept was selected after it became apparent that a conventional power generator could not satisfy the ML-1 weight and size requirements. The demonstration alternator fabricated and tested under this program did not meet all performance specifications; the efficiency was low and the unit could not be operated for significant periods of time without overheating. However, a large body of useful data was accumulated during the extensive development program. Of special interest are data on the rotor and stator design, the cooling requirements and on the distribution of eddy current losses. Analysis of the data indicates that a brushless alternator, only slightly larger and heavier than was specified for the ML-1, could be developed with a modest additional effort.

  8. Gas-cooled reactor commercialization study: introduction scenario and commercialization analyses for process heat applications. Final report, July 8, 1977--November 30, 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-12-01

    This report identifies and presents an introduction scenario which can lead to the operation of High Temperature Gas Cooled Reactor demonstration plants for combined process heat and electric power generation applications, and presents a commercialization analysis relevant to the organizational and management plans which could implement a development program.

  9. TINTE. Nuclear calculation theory description report

    Energy Technology Data Exchange (ETDEWEB)

    Gerwin, H.; Scherer, W.; Lauer, A. [Forschungszentrum Juelich GmbH (DE). Institut fuer Energieforschung (IEF), Sicherheitsforschung und Reaktortechnik (IEF-6); Clifford, I. [Pebble Bed Modular Reactor (Pty) Ltd. (South Africa)

    2010-01-15

    The Time Dependent Neutronics and Temperatures (TINTE) code system deals with the nuclear and the thermal transient behaviour of the primary circuit of the High-temperature Gas-cooled Reactor (HTGR), taking into consideration the mutual feedback effects in twodimensional axisymmetric geometry. This document contains a complete description of the theoretical basis of the TINTE nuclear calculation, including the equations solved, solution methods and the nuclear data used in the solution. (orig.)

  10. Advances in processing technologies for titanium heat exchanger tubes of fossil and nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Likhareva, T.P.; Tchizhik, A.A.; Chavchanidze, N.N. [Polzanov Central Boiler and Turbine Institute, St. Petersburg (Russian Federation)

    1998-12-31

    The advances in processing technologies for titanium heat exchangers with rolled and welded tubes of fossil and nuclear power plants in Russia are presented. The special methodology of investigations with constant small strain rate have been used to study the effects of mixed corrosion and creep processes in condensers cooled by sea or synthetic sea waters. The results of corrosion creep tests and K1scc calculations are given. The Russian science activities concerning condensers manufactured from titanium show the possibilities for designing structures with very high level service reliability in different corrosion aggressive mediums with high total salt, Cl-ion and oxygen contents. (orig.)

  11. Advanced high brightness ion rf accelerator applications in the nuclear energy

    Energy Technology Data Exchange (ETDEWEB)

    Jameson, R.A.

    1991-01-01

    The capability of modern rf linear accelerators to provide intense high quality beams of protons, deuterons, or heavier ions is opening new possibilities for transmuting existing nuclear wastes, for generating electricity from readily available fuels with minimal residual wastes, for building intense neutron sources for materials research, for inertial confinement fusion using heavy ions, and for other new applications. These are briefly described, couched in a perspective of the advances in the understanding of the high brightness beams that has enabled these new programs. 32 refs., 2 figs.

  12. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    Energy Technology Data Exchange (ETDEWEB)

    Van Hecke, K.; Goethals, P.

    2006-07-15

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  13. UNEDF: Advanced Scientific Computing Collaboration Transforms the Low-Energy Nuclear Many-Body Problem

    CERN Document Server

    Nam, H; Nazarewicz, W; Bulgac, A; Hagen, G; Kortelainen, M; Maris, P; Pei, J C; Roche, K J; Schunck, N; Thompson, I; Vary, J P; Wild, S M

    2012-01-01

    The demands of cutting-edge science are driving the need for larger and faster computing resources. With the rapidly growing scale of computing systems and the prospect of technologically disruptive architectures to meet these needs, scientists face the challenge of effectively using complex computational resources to advance scientific discovery. Multidisciplinary collaborating networks of researchers with diverse scientific backgrounds are needed to address these complex challenges. The UNEDF SciDAC collaboration of nuclear theorists, applied mathematicians, and computer scientists is developing a comprehensive description of nuclei and their reactions that delivers maximum predictive power with quantified uncertainties. This paper describes UNEDF and identifies attributes that classify it as a successful computational collaboration. We illustrate significant milestones accomplished by UNEDF through integrative solutions using the most reliable theoretical approaches, most advanced algorithms, and leadershi...

  14. The Nuclear Energy Advanced Modeling and Simulation Enabling Computational Technologies FY09 Report

    Energy Technology Data Exchange (ETDEWEB)

    Diachin, L F; Garaizar, F X; Henson, V E; Pope, G

    2009-10-12

    In this document we report on the status of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Enabling Computational Technologies (ECT) effort. In particular, we provide the context for ECT In the broader NEAMS program and describe the three pillars of the ECT effort, namely, (1) tools and libraries, (2) software quality assurance, and (3) computational facility (computers, storage, etc) needs. We report on our FY09 deliverables to determine the needs of the integrated performance and safety codes (IPSCs) in these three areas and lay out the general plan for software quality assurance to meet the requirements of DOE and the DOE Advanced Fuel Cycle Initiative (AFCI). We conclude with a brief description of our interactions with the Idaho National Laboratory computer center to determine what is needed to expand their role as a NEAMS user facility.

  15. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, April 1, 1983-September 30, 1983

    Energy Technology Data Exchange (ETDEWEB)

    1983-12-01

    An assessment of the HTGR opportunities from the year 2000 through 2045 was the principal activity on the Market Definition Task (WBS 03). Within the Plant Technology (WBS 13) task, there were activities to develop analytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. The activities in support of the HTGR-SC/C Lead Plant (WBS 30 and 31) were the participation in the Lead Plant System Engineering (LPSE) effort and the plant simulation task. The efforts on the Advanced HTGR systems was performed under the Modular Reactor Systems (MRS) (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors.

  16. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  17. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

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

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

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

  19. 76 FR 48184 - Exelon Nuclear, Peach Bottom Atomic Power Station, Unit 1; Exemption From Certain Security...

    Science.gov (United States)

    2011-08-08

    ... a permanently shut down nuclear reactor facility. PBAPS Unit 1 was a high-temperature, gas-cooled... nuclear power reactors against radiological sabotage,'' paragraph (b)(1) states, ``The licensee shall... its objective to provide high assurance that activities involving special nuclear material are...

  20. The development of advanced robotics for the nuclear industry -The development of robotic system for the nuclear power plants-

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Hoh; Kim, Chang Heui; Kim, Byung Soo; Hwang, Suk Yong; Kim, Kee Hoh; Lee, Jae Kyung; Lee, Yung Kwang; Suh, Yong Chil; Lee, Yong Bum; Kim, Woong Kee; Park, Soon Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The Advanced Robotics Department of the Korea Atomic Energy Research Institute (KAERI) is developing a Dexterous Manipulation System (DMS) and a teleoperated mobile robot, identified as KAEROT/ml, for use in nuclear power plants. The DMS is being developed for performing tasks such as the opening and closing of nozzle dam inside water chamber of steam generator. The DMS has two major subsystems; a master-slave 6 degrees of freedom (dof) manipulator and a support device. The master-slave arms are designed dissimilar kinematically and dynamically, and their functions are performed by a bilateral force-reflecting force control. The slave is a hydraulically powered arm with a 3 dof end effector, and is mounted on the top of the support device for nozzle dam operation. The support device guides the slave arm into the water chamber and supports it during its operation. The DMS can be operated either in teleoperated or supervisory control modes. The KAEROT/ml is designed to be used in emergency response applications such as monitoring and mapping radiation areas, handling radioactive materials and performing decontamination tasks. The KAEROT/ml equipped with four-omnidirectional planetary wheels has a 6 dof joint-controlled arm and is capable of ascending and descending stairs and navigating flat surface with zero turning radius. This report describes the mechanical design, features, modeling and control system of both the DMS and the KAEROT/ml. 209 figs, 49 pix, 69 tabs, 62 refs. (Author).

  1. Assessment of modular construction for safety-related structures at advanced nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Braverman, J.; Morante, R.; Hofmayer, C.

    1997-03-01

    Modular construction techniques have been successfully used in a number of industries, both domestically and internationally. Recently, the use of structural modules has been proposed for advanced nuclear power plants. The objective in utilizing modular construction is to reduce the construction schedule, reduce construction costs, and improve the quality of construction. This report documents the results of a program which evaluated the proposed use of modular construction for safety-related structures in advanced nuclear power plant designs. The program included review of current modular construction technology, development of licensing review criteria for modular construction, and initial validation of currently available analytical techniques applied to concrete-filled steel structural modules. The program was conducted in three phases. The objective of the first phase was to identify the technical issues and the need for further study in order to support NRC licensing review activities. The two key findings were the need for supplementary review criteria to augment the Standard Review Plan and the need for verified design/analysis methodology for unique types of modules, such as the concrete-filled steel module. In the second phase of this program, Modular Construction Review Criteria were developed to provide guidance for licensing reviews. In the third phase, an analysis effort was conducted to determine if currently available finite element analysis techniques can be used to predict the response of concrete-filled steel modules.

  2. Development of Advanced Nuclide Separation and Recovery Methods using Ion-Exchanhge Techniques in Nuclear Backend

    Science.gov (United States)

    Miura, Hitoshi

    The development of compact separation and recovery methods using selective ion-exchange techniques is very important for the reprocessing and high-level liquid wastes (HLLWs) treatment in the nuclear backend field. The selective nuclide separation techniques are effective for the volume reduction of wastes and the utilization of valuable nuclides, and expected for the construction of advanced nuclear fuel cycle system and the rationalization of waste treatment. In order to accomplish the selective nuclide separation, the design and synthesis of novel adsorbents are essential for the development of compact and precise separation processes. The present paper deals with the preparation of highly functional and selective hybrid microcapsules enclosing nano-adsorbents in the alginate gel polymer matrices by sol-gel methods, their characterization and the clarification of selective adsorption properties by batch and column methods. The selective separation of Cs, Pd and Re in real HLLW was further accomplished by using novel microcapsules, and an advanced nuclide separation system was proposed by the combination of selective processes using microcapsules.

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

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

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

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

    Science.gov (United States)

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

    2011-01-01

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

  5. Development of an evaluation method of fission product release fraction from High Temperature Gas-cooled Reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sawa, Kazuhiro; Minato, Kazuo; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-11-01

    The High Temperature Gas-cooled Reactor (HTGR) uses coated particles as fuel. Current coated particle is a microsphere of fuel kernel with TRISO coatings. The TRISO coatings consist of a low-density, porous pyrolytic carbon (PyC) buffer layer adjacent to the spherical fuel kernel, followed by an inner isotropic PyC layer, a SiC layer and a final (outer) PyC layer. An evaluation method of fission product release behavior during the normal operation was developed. Key issues of fission gas release model were: (1) fission gas releases from matrix contamination uranium and through-coatings failed particle were separately modeled and (2) burnup and fast neutron irradiation effects were newly considered. For metallic fission product, fractional release of cesium from coated fuel particles was investigated by comparing measured data in an irradiation test which contained three kinds of fuel particles; artificially bored particles simulating through-coatings failed particles, as-manufactured SiC-failed particles and intact particles. Through the comparison of measured and calculated fractional releases, an equivalent diffusion coefficient of SiC layer in the SiC-failed particle was introduced. This report describes the developed model together with validation result of the release model. (author)

  6. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  7. Efficient transfer of positrons from a buffer-gas-cooled accumulator into an orthogonally oriented superconducting solenoid for antihydrogen studies

    CERN Document Server

    Comeau, D; Fitzakerley, D; George, M C; Hessels, E A; Storry, C H; Weel, M; Grzonka, D; Oelert, W; Gabrielse, G; Kalra, R; Kolthammer, W S; McConnell, R; Richerme, P; Mullers, A; Walz, J

    2012-01-01

    Positrons accumulated in a room-temperature buffer-gas-cooled positron accumulator are efficiently transferred into a superconducting solenoid which houses the ATRAP cryogenic Penning trap used in antihydrogen research. The positrons are guided along a 9 m long magnetic guide that connects the central field lines of the 0.15 T field in the positron accumulator to the central magnetic field lines of the superconducting solenoid. Seventy independently controllable electromagnets are required to overcome the fringing field of the large-bore superconducting solenoid. The guide includes both a 15 degrees upward bend and a 105 degrees downward bend to account for the orthogonal orientation of the positron accumulator with respect to the cryogenic Penning trap. Low-energy positrons ejected from the accumulator follow the magnetic field lines within the guide and are transferred into the superconducting solenoid with nearly 100% efficiency. A 7 m long 5 cm diameter stainless-steel tube and a 20 mm long, 1.5 mm diamet...

  8. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  9. Detection of hidden shot balls in a gas-cooled turbine blade with an NRT gadolinium tagging method

    Science.gov (United States)

    Sim, Cheul Muu; Kim, Yi Kyung; Kim, TaeJoo; Lee, Kye Hong; Kim, Jeong Uk

    2009-06-01

    This report provides a preliminary insight into the benefits and effectiveness of neutron radiography in identifying alien materials, namely shot balls hidden in a turbine blade that are otherwise undetected using other methods. The detection of 0.2-mm-diameter shot balls in gas-cooled turbine blades is possible for thermal neutron radiography. A tagging processing is more useful for a distinctive image of newer turbine blades. Areas of concern for the tagging process include the solution concentration and the possibility of a slight washing of the blades. The location of the shot balls within the turbine blades tagged with Gd((2%, 5%)+water) was shown. Shot balls were placed externally on a turbine blade (F100-700, F100-200) surface in order to check for a dead zone from a surface examination. The image is produced from neutron radiography after a 5 min exposure time. When the blade is tagged with 2% and 5% Gd with slight washing, the shot can also be effectively seen on the SR-45 film. Shot balls are more obvious on a neutron image SR-45 film than an image plate or a DR film.

  10. Model based predictive control of a high temperature gas cooled power plant coupled to a hydrogen production facility

    Science.gov (United States)

    Rhoads, Lloyd A.

    This thesis builds upon recent studies focusing on modeling, operation, and control of high temperature gas cooled reactors. A computer model was developed, based on mass, energy, and momentum balances of control volumes throughout the plant. Several simulations of the plant behavior were conducted and their results were compared with those from the literature. Proportional control was combined with optimal control to form a time varying, adjustable gain predictive controller which adjusts the proportional gains during transients. The controller was designed to utilize control rod motions and bypass control valves to maintain desired plant conditions. An optimization scheme was introduced to efficiently solve the optimization problem formulated as part of the predictive controller operation. Several additional transients were run to examine the full plant controller performance. Multiple predictive controllers were designed and their performance was compared with a proportional controller throughout each transient. The predictive controller results confirmed the importance of proper selection of the optimal controller parameters, in particular the controller time step size and the horizon time. The well-designed proportional controllers clearly demonstrated improvements in plant performance during short time scale transients, namely a loss of secondary heat transfer transient and a step change in desired power transient. Results from long time scale transients demonstrated the capabilities of the proposed bypass control system to control electrical power production without the need for storage vessels.

  11. Design Study of 200MWth Gas Cooled Fast Reactor with Nitride (UN-PuN Fuel Long Life without Refueling

    Directory of Open Access Journals (Sweden)

    Syarifah Ratna Dewi

    2016-01-01

    Full Text Available Design study of 200 MWth Gas Cooled Fast Reactor with UN-PuN fuel long life without refueling has been done. GFR is one type reactor in Generation IV reactor system. It uses helium coolant and fast neutron spectrum. Helium is chemical inert, single phase and low neutron moderation. In this study the calculations are performed by using SRAC code with PIJ calculation for the fuel pin cell calculation and CITATION calculation for core calculation. The data libraries use JENDL 3.2. The variation fuel fractions are 50% until 60%. The diameter active core is 150 cm and the height active core is 100 cm. The reflector radial-axial width is 50 cm. The variation of the powers are 100 MWth up to 500 MWth. The high power causes the high k-eff value. The optimum design is reached when the power is 200 MWth, variation percentage Plutonium for fuel F1:F2:F3=9%:11%:13%. The comparation of fuel:cladding:coolant fraction = 55%:10%:35%. The cooling down time of Plutonium is nine months. The optimum k-eff value is 1.0142 with excess reactivity value 1.403%. The decay of Plutonium decrease k-eff value in the beginning of burn up.

  12. Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

    Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

  13. High Level Requirements for the Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

    Energy Technology Data Exchange (ETDEWEB)

    Rich Johnson; Hyung Lee; Kimberlyn C. Mousseau

    2011-09-01

    The US Department of Energy, Office of Nuclear Energy (DOE-NE), has been tasked with the important mission of ensuring that nuclear energy remains a compelling and viable energy source in the U.S. The motivations behind this mission include cost-effectively meeting the expected increases in the power needs of the country, reducing carbon emissions and reducing dependence on foreign energy sources. In the near term, to ensure that nuclear power remains a key element of U.S. energy strategy and portfolio, the DOE-NE will be working with the nuclear industry to support safe and efficient operations of existing nuclear power plants. In the long term, to meet the increasing energy needs of the U.S., the DOE-NE will be investing in research and development (R&D) and working in concert with the nuclear industry to build and deploy new, safer and more efficient nuclear power plants. The safe and efficient operations of existing nuclear power plants and designing, licensing and deploying new reactor designs, however, will require focused R&D programs as well as the extensive use and leveraging of advanced modeling and simulation (M&S). M&S will play a key role in ensuring safe and efficient operations of existing and new nuclear reactors. The DOE-NE has been actively developing and promoting the use of advanced M&S in reactor design and analysis through its R&D programs, e.g., the Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Consortium for Advanced Simulation of Light Water Reactors (CASL) programs. Also, nuclear reactor vendors are already using CFD and CSM, for design, analysis, and licensing. However, these M&S tools cannot be used with confidence for nuclear reactor applications unless accompanied and supported by verification and validation (V&V) and uncertainty quantification (UQ) processes and procedures which provide quantitative measures of uncertainty for specific applications. The Nuclear Energy Knowledge base for Advanced Modeling and Simulation

  14. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  15. Development of computational methods for the safety assessment of gas-cooled high-temperature and supercritical light-water reactors. Final report; Rechenmethoden zur Bewertung der Sicherheit von gasgekuehlten Hochtemperaturreaktoren und superkritischen Leichtwasserreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, S.; Cron, D. von der; Hristov, H.; Lerchl, G.; Papukchiev, A.; Seubert, A.; Sureda, A.; Weis, J.; Weyermann, F.

    2012-12-15

    This report documents developments and results in the frame of the project RS1191 ''Development of computational methods for the safety assessment of gas-cooled high temperature and supercritical light-water reactors''. The report is structured according to the five work packages: 1. Reactor physics modeling of gas-cooled high temperature reactors; 2. Coupling of reactor physics and 3-D thermal hydraulics for the core barrel; 3. Extension of ATHLET models for application to supercritical reactors (HPLWR); 4. Further development of ATHLET for application to HTR; 5. Further development and validation of ANSYS CFX for application to alternative reactor concepts. Chapter 4 describes the extensions made in TORT-TD related to the simulation of pebble-bed HTR, e.g. spectral zone buckling, Iodine-Xenon dynamics, nuclear decay heat calculation and extension of the cross section interpolation algorithms to higher dimensions. For fast running scoping calculations, a time-dependent 3-D diffusion solver has been implemented in TORT-TD. For the PBMR-268 and PBMR-400 as well as for the HTR-10 reactor, appropriate TORT-TD models have been developed. Few-group nuclear cross sections have been generated using the spectral codes MICROX- 2 and DRAGON4. For verification and validation of nuclear cross sections and deterministic reactor models, MCNP models of reactor core and control rod of the HTR-10 have been developed. Comparisons with experimental data have been performed for the HTR-10 first criticality and control rod worth. The development of the coupled 3-D neutron kinetics and thermal hydraulics code system TORT-TD/ATTICA3D is documented in chapter 5. Similar to the couplings with ATHLET and COBRA-TF, the ''internal'' coupling approach has been implemented. Regarding the review of experiments and benchmarks relevant to HTR for validation of the coupled code system, the PBMR-400 benchmarks and the HTR-10 test reactor have been selected

  16. Progress in high energy physics and nuclear safety : Proceedings of the NATO Advanced Research Workshop on Safe Nuclear Energy

    CERN Document Server

    Polański, Aleksander; Begun, Viktor

    2009-01-01

    The book contains recent results on the progress in high-energy physics, accelerator, detection and nuclear technologies, as well as nuclear safety in high-energy experimentation and in nuclear industry, covered by leading experts in the field. The forthcoming experiments at the Large Hadron Collider (LHC) at CERN and cosmic-ray experiments are highlighted. Most of the current high-energy experiments and their physical motivation are analyzed. Various nuclear energy safety aspects, including progress in the production of new radiation-resistant materials, new and safe nuclear reactor designs, such as the slowly-burning reactor, as well as the use of coal-nuclear symbiotic methods of energy production can be found in the book.

  17. Recent advances in nuclear fission theory: pre- and post-scission physics

    Energy Technology Data Exchange (ETDEWEB)

    Talou, Patrick [Los Alamos National Laboratory; Kawano, Toshihiko [Los Alamos National Laboratory; Bouland, Olivier [Los Alamos National Laboratory; Moller, Peter [Los Alamos National Laboratory; Chadwick, Mark B [Los Alamos National Laboratory; Lynn, J E [CEA DEN FRANCE

    2010-01-01

    Recent advances in the modeling of the nuclear fission process for data evaluation purposes are reviewed. In particular, it is stressed that a more comprehensive approach to fission data is needed if predictive capability is to be achieved. The link between pre- and post-scission data is clarified, and a path forward to evaluate those data in a consistent and comprehensive manner is presented. Two examples are given: (i) the modeling of fission cross-sections in the R-matrix formalism, for which results for Pu isotopes from 239 to 242 are presented; (ii) the modeling of prompt fission neutrons in the Monte Carlo Hauser-Feshbach framework. Results for neutron-induced fission on {sup 235}U are discussed.

  18. Intelligent software system for the advanced control room of a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Choi, Seong Soo; Park, Jin Kyun; Heo, Gyung Young [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Han Gon [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The intelligent software system for nuclear power plants (NPPs) has been conceptually designed in this study. Its design goals are to operate NPPs in an improved manner and to support operators` cognitive takes. It consists of six major modules such as {sup I}nformation Processing,{sup {sup A}}larm Processing,{sup {sup P}}rocedure Tracking,{sup {sup P}}erformance Diagnosis,{sup a}nd {sup E}vent Diagnosis{sup m}odules for operators and {sup M}alfunction Diagnosis{sup m}odule for maintenance personnel. Most of the modules have been developed for several years and the others are under development. After the completion of development, they will be combined into one system that would be main parts of advanced control rooms in NPPs. 5 refs., 4 figs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

  20. Uncertainty quantification approaches for advanced reactor analyses.

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

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

  1. Advances and applications of dynamic-angle spinning nuclear magnetic resonance

    Energy Technology Data Exchange (ETDEWEB)

    Baltisberger, J.H.

    1993-06-01

    This dissertation describes nuclear magnetic resonance experiments and theory which have been developed to study quadrupolar nuclei (those nuclei with spin greater than one-half) in the solid state. Primarily, the technique of dynamic-angle spinning (DAS) is extensively reviewed and expanded upon in this thesis. Specifically, the improvement in both the resolution (two-dimensional pure-absorptive phase methods and DAS angle choice) and sensitivity (pulse-sequence development), along with effective spinning speed enhancement (again through choice of DAS conditions or alternative multiple pulse schemes) of dynamic-angle spinning experiment was realized with both theory and experimental examples. The application of DAS to new types of nuclei (specifically the {sup 87}Rb and {sup 85}Rb nuclear spins) and materials (specifically amorphous solids) has also greatly expanded the possibilities of the use of DAS to study a larger range of materials. This dissertation is meant to demonstrate both recent advances and applications of the DAS technique, and by no means represents a comprehensive study of any particular chemical problem.

  2. 球床式高温气冷堆的余热不确定性分析%Uncertainty Analysis on Decay Heat of Pebble-bed High Temperature Gas-cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    贠相羽; 郑艳华; 经荥清; 李富

    2013-01-01

    反应堆在停堆后相当长时间内仍具有较高的剩余发热是核电站的重要特性,也是核电站安全分析的关键.因此,对反应堆余热及其不确定性进行分析,对于合理设计余热排出系统、研究论证燃料元件在事故后的安全特性等均具有重要意义.本工作结合德国针对球床式高温气冷堆制定的余热计算标准,介绍了球床式高温气冷堆剩余发热及其不确定性的计算方法,并结合200 MWe球床模块式高温气冷堆示范工程(HTR-PM)的初步物理设计,对长期运行在满功率平衡堆芯状态下的反应堆停堆后的余热及其不确定性进行了计算分析,为进一步的事故分析提供依据.%The large amount of decay heat in a quite long time after reactor shutdown,which is an important characteristic of the nuclear power plants,should be considered seriously during the safety analysis.Therefore,the study on the decay heat and its uncertainty analysis play an important role in the design of decay heat removal system,as well as in the safety verification of the fuel element during the accident.In referenced to the standard of Germany entitled "Decay Heat Power in Nuclear Fuels of Hightemperature Reactors with Spherical Fuel Elements" especially for pebble-bed high temperature gas-cooled reactor (HTGR),the calculation method of decay heat and its uncertainty of pebble-bed HTGR were introduced.On the basis of the preliminary physical design of Chinese 200 MWe high temperature gas-cooled reactor pebble-bed module (HTR-PM),the decay heat and its uncertainty after reactor shutdown from long-term operation at rated power were analyzed,so as to provide a basis for further accident analysis.

  3. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Directory of Open Access Journals (Sweden)

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

  4. Reliability study of a special decay heat removal system of a gas-cooled fast reactor demonstrator

    Energy Technology Data Exchange (ETDEWEB)

    Burgazzi, Luciano, E-mail: luciano.burgazzi@enea.it

    2014-12-15

    The European roadmap toward the development of generation IV concepts addresses the safety and reliability assessment of the special system designed for decay heat removal of a gas-cooled fast reactor demonstrator (GFRD). The envisaged system includes the combination of both active and passive means to accomplish the fundamental safety function. Failure probabilities are calculated on various system configurations, according to either pressurized or depressurized accident events under investigation, and integrated with probabilities of occurrence of corresponding hardware components and natural circulation performance assessment. The analysis suggests the improvement of measures against common cause failures (CCF), in terms of an appropriate diversification among the redundant systems, to reduce the system failure risk. Particular emphasis is placed upon passive system reliability assessment, being recognized to be still an open issue, and the approach based on the functional reliability is adopted to address the point. Results highlight natural circulation as a challenging factor for the decay heat removal safety function accomplishment by means of passive devices. With the models presented here, the simplifying assumptions and the limited scenarios considered according to the level of definition of the design, where many systems are not yet established, one can conclude that attention has to be paid to the functional aspects of the passive system, i.e. the ones not pertaining to the “hardware” of the system. In this article the results of the analysis are discussed, where the effects of the analytical assumptions, design options, accident managements on the reliability are examined. The design diversity of the components undergoing CCFs can be effective for the improvement and some accident management measures are also possible by making use of the long grace period in GFRD.

  5. Recent Advances in Characterization of Lignin Polymer by Solution-State Nuclear Magnetic Resonance (NMR Methodology

    Directory of Open Access Journals (Sweden)

    Run-Cang Sun

    2013-01-01

    Full Text Available The demand for efficient utilization of biomass induces a detailed analysis of the fundamental chemical structures of biomass, especially the complex structures of lignin polymers, which have long been recognized for their negative impact on biorefinery. Traditionally, it has been attempted to reveal the complicated and heterogeneous structure of lignin by a series of chemical analyses, such as thioacidolysis (TA, nitrobenzene oxidation (NBO, and derivatization followed by reductive cleavage (DFRC. Recent advances in nuclear magnetic resonance (NMR technology undoubtedly have made solution-state NMR become the most widely used technique in structural characterization of lignin due to its versatility in illustrating structural features and structural transformations of lignin polymers. As one of the most promising diagnostic tools, NMR provides unambiguous evidence for specific structures as well as quantitative structural information. The recent advances in two-dimensional solution-state NMR techniques for structural analysis of lignin in isolated and whole cell wall states (in situ, as well as their applications are reviewed.

  6. Recent advances in medical device triage technologies for chemical, biological, radiological, and nuclear events.

    Science.gov (United States)

    Lansdowne, Krystal; Scully, Christopher G; Galeotti, Loriano; Schwartz, Suzanne; Marcozzi, David; Strauss, David G

    2015-06-01

    In 2010, the US Food and Drug Administration (Silver Spring, Maryland USA) created the Medical Countermeasures Initiative with the mission of development and promoting medical countermeasures that would be needed to protect the nation from identified, high-priority chemical, biological, radiological, or nuclear (CBRN) threats and emerging infectious diseases. The aim of this review was to promote regulatory science research of medical devices and to analyze how the devices can be employed in different CBRN scenarios. Triage in CBRN scenarios presents unique challenges for first responders because the effects of CBRN agents and the clinical presentations of casualties at each triage stage can vary. The uniqueness of a CBRN event can render standard patient monitoring medical device and conventional triage algorithms ineffective. Despite the challenges, there have been recent advances in CBRN triage technology that include: novel technologies; mobile medical applications ("medical apps") for CBRN disasters; electronic triage tags, such as eTriage; diagnostic field devices, such as the Joint Biological Agent Identification System; and decision support systems, such as the Chemical Hazards Emergency Medical Management Intelligent Syndromes Tool (CHEMM-IST). Further research and medical device validation can help to advance prehospital triage technology for CBRN events.

  7. The sixth Conference on Advanced Topics in the Interdisciplinary Fields of Particle Physics, Nuclear Physics and Cosmology

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    @@ The sixth Conference on Advanced Topics in the Interdisciplinary Fields of Particle Paysics.Nuclear Paysic8and Cosmology was held on July 22-27,2007 in Lijiang,Yunan Province.Over 70 scientists who were activelv working in the fields attended the conference and 42 physicists gave talks.9 papers have been selected to be published in the proceedings.

  8. Nuclear Data Needs for Generation IV Nuclear Energy Systems

    Science.gov (United States)

    Rullhusen, Peter

    2006-04-01

    Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S

  9. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is

  10. Numerical Simulation of Plasma-Dynamical Processes in the Technological Inductively Coupled RF Plasmatron with Gas Cooling

    Directory of Open Access Journals (Sweden)

    Yu. M. Grishin

    2016-01-01

    Full Text Available The electrodeless inductively coupled RF plasmatron (ICP torches became widely used in various fields of engineering, science and technology. Presently, owing to development of new technologies to produce very pure substances, nanopowders, etc., there is a steadily increasing interest in the induction plasma. This generates a need for a broad range of theoretical and experimental studies to optimize the design and technological parameters of different ICP equipment.The paper presents a numerical model to calculate parameters of inductively coupled RF plasmatron with gas-cooling flow. A finite volume method is used for a numerical solution of a system of Maxwell's and heat transfer equations in the application package ANSYS CFX (14.5. The pseudo-steady approach to solving problems is used.A numerical simulation has been computed in the application package ANSYS CFX (14.5 for a specific design option of the technological ICP, which has a three-coils inductor and current amplitude in the range J к = 50-170 A (3 MHz. The pure argon flows in the ICP. The paper discusses how the value of discharge current impacts on the thermodynamic parameters (pressure, temperature and the power energy in discharge zone. It shows that the ICP can generate a plasma stream with a maximum temperature of about 10 kK and an output speed of 10-15 m/s. The work determines a length of the plasma stream with a weight average temperature of more than 4 kK. It has been found that in order to keep the quartz walls in normal thermal state, the discharge current amplitude should not exceed 150 A. The paper shows the features of the velocity field distribution in the channel of the plasma torch, namely, the formation of vortex in the position of the first coil. The results obtained are important for calculating the dynamics of heating and evaporation of quartz particles in the manufacturing processes for plasma processing of quartz concentrate into high-purity quartz and

  11. 基于微机系统的高温气冷堆工程仿真机%A Personal Computer-Based Engineering Simulator for High Temperature Gas-Cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    石磊; 高祖瑛

    2001-01-01

    A personal computer(PC)-based engineering simulator for high temperature gas-cooled reactor(HTR), HTRSIMU has been developed by Institute of Nuclear Energy Technology(INET) of Tsinghua University. HTRSIMU runs on Windows 98 or Windows 2000 operating system, in which the multi-process and multi-monitor techniques are applied.The graphical user interface is friendly and easy to operate.The basic models include the reactor core, the main loop system and steam generator of 10MW high temperature gas-cooled reactor(HTR-10), which can be simulated detailed in normal operation and transient accidents.The simulation results are in good agreement with design data and the safety analysis reports of HTR-10.HTRSIMU can be used as an education al tool to understand the design and operational characteristics of a HTR. In the future, HTRSIMU will provide on -line supports for operators in the main control room of HTR-10%基于微机系统的高温气冷堆工程仿真机 (HTRSIMU)由清华大学核能技术设计研究院开发完成。 HTRSIMU运行于 Windows98或 Windows2000平台上,采用多进程、多显示器结构,具有人机界面友好、结构紧凑、操作方便、易于扩展等特点。它的模型包括 10MW高温气冷堆 (HTR-10)的一、二回路主要部件,能够对反应堆堆芯、主回路系统和蒸汽发生器等部件做详细的物理和热工分析计算,可以模拟正常运行和各种事故工况过程,仿真结果与 HTR-10的设计值和安全分析报告符合得很好。利用 HTRSIMU系统不仅可以进行高温气冷堆的工程设计、安全分析和人员培训,而且将来可以对 HTR-10主控室的操纵人员进行现场支持,给实际运行和各项研究提供帮助。

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-01

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

  13. 10 MW高温气冷堆反应堆压力容器的出厂水压试验%Hydraulic Pressure Test of Pressure Vessel of 10 MW High Temperature Gas-cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    刘俊杰; 张征明; 何树延; 王金海

    2001-01-01

    The hydraulic pressure test of 10MW Hight Temperature Gas-cooled Reactorc(HTR-10) pressure vessel was successfully performed according to the requirement of the section NB-6200, ASME Ⅲ code. The test requirement, the test results and the test evaluations are described in detail. The test tension was effectively and rationally done through an hydraulic tensionor, which was developed at institue of nuclear energy technology of Tsinghua University. The strain and deformation of the HTR-10 pressure vessel were also measured.%根据ASME规范第Ⅲ卷NB-6200节的规定,对10MW高温气冷堆压力容器的水压试验要求、试验过程,试验结果及评价进行了叙述。用清华大学核能技术设计研究院研制的液压张拉机对主螺栓实施了合理及有效的张拉,对压力容器进行了应变和变形测量,取得了反应堆压力容器水压试验的圆满成功。

  14. Advancing the Theory of Nuclear Reactions with Rare Isotopes: From the Laboratory to the Cosmos

    Energy Technology Data Exchange (ETDEWEB)

    Elster, Charlotte [Ohio Univ., Athens, OH (United States)

    2015-06-01

    The mission of the TORUS Topical Collaboration is to develop new methods that will advance nuclear reaction theory for unstable isotopes by using three-body techniques to improve direct-reaction calculations, and, by using a new partial-fusion theory, to integrate descriptions of direct and compound-nucleus reactions. Ohio University concentrates its efforts on the first part of the mission. Since direct measurements are often not feasible, indirect methods, e.g. (d,p) reactions, should be used. Those (d,p) reactions may be viewed as three-body reactions and described with Faddeev techniques. Faddeev equations in momentum space have a long tradition of utilizing separable interactions in order to arrive at sets of coupled integral equations in one variable. While there exist several separable representations for the nucleon-nucleon interaction, the optical potential between a neutron (proton) and a nucleus is not readily available in separable form. For this reason we first embarked in introducing a separable representation for complex phenomenological optical potentials of Woods-Saxon type.

  15. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, Peter Andrew

    2011-12-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  16. Advancing the Theory of Nuclear Reactions with Rare Isotopes. From the Laboratory to the Cosmos

    Energy Technology Data Exchange (ETDEWEB)

    Nunes, Filomena [Michigan State Univ., East Lansing, MI (United States)

    2015-06-01

    The mission of the Topical Collaboration on the Theory of Reactions for Unstable iSotopes (TORUS) was to develop new methods to advance nuclear reaction theory for unstable isotopes—particularly the (d,p) reaction in which a deuteron, composed of a proton and a neutron, transfers its neutron to an unstable nucleus. After benchmarking the state-of-the-art theories, the TORUS collaboration found that there were no exact methods to study (d,p) reactions involving heavy targets; the difficulty arising from the long-range nature of the well known, yet subtle, Coulomb force. To overcome this challenge, the TORUS collaboration developed a new theory where the complexity of treating the long-range Coulomb interaction is shifted to the calculation of so-called form-factors. An efficient implementation for the computation of these form factors was a major achievement of the TORUS collaboration. All the new machinery developed are essential ingredients to analyse (d,p) reactions involving heavy nuclei relevant for astrophysics, energy production, and stockpile stewardship.

  17. Advance: research project on aging electrical wiring in nuclear power plants; Advance: proyecto de investigacion de envejecimiento en cableado electrico en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Cano, J. C.; Ruiz, S.

    2013-07-01

    As Nuclear Power Plants get older it is more important to know the real condition of low voltage, instrumentation, power and control cables. Additionally, as new plants are being built, the election of cables and the use of in-situ monitoring techniques to get reliable aging indicators, can be very useful during the plant life. The goal of this Project is to adapt, optimize and asses Condition Monitoring techniques for Nuclear Power Plants cables. These techniques, together with the appropriate acceptance criteria, would allow specialists to know the state of the cable over its entire length and estimate its residual life. In the Project, accelerated ageing is used in cables installed in European NPPs in order to evaluate different techniques to detect local and global ageing. Results are compared with accepted tests to validate its use for the estimation of cables residual life. This paper describes the main stages of the Project and some results. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  19. Analysis of advanced european nuclear fuel cycle scenarios including transmutation and economical estimates

    Energy Technology Data Exchange (ETDEWEB)

    Merino Rodriguez, I.; Alvarez-Velarde, F.; Martin-Fuertes, F. [CIEMAT, Avda. Complutense, 40, 28040 Madrid (Spain)

    2013-07-01

    In this work the transition from the existing Light Water Reactors (LWR) to the advanced reactors is analyzed, including Generation III+ reactors in a European framework. Four European fuel cycle scenarios involving transmutation options have been addressed. The first scenario (i.e., reference) is the current fleet using LWR technology and open fuel cycle. The second scenario assumes a full replacement of the initial fleet with Fast Reactors (FR) burning U-Pu MOX fuel. The third scenario is a modification of the second one introducing Minor Actinide (MA) transmutation in a fraction of the FR fleet. Finally, in the fourth scenario, the LWR fleet is replaced using FR with MOX fuel as well as Accelerator Driven Systems (ADS) for MA transmutation. All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for a period of 200 years looking for equilibrium mass flows. The simulations were made using the TR-EVOL code, a tool for fuel cycle studies developed by CIEMAT. The results reveal that all scenarios are feasible according to nuclear resources demand (U and Pu). Concerning to no transmutation cases, the second scenario reduces considerably the Pu inventory in repositories compared to the reference scenario, although the MA inventory increases. The transmutation scenarios show that elimination of the LWR MA legacy requires on one hand a maximum of 33% fraction (i.e., a peak value of 26 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation). On the other hand a maximum number of ADS plants accounting for 5% of electricity generation are predicted in the fourth scenario (i.e., 35 ADS units). Regarding the economic analysis, the estimations show an increase of LCOE (Levelized cost of electricity) - averaged over the whole period - with respect to the reference scenario of 21% and 29% for FR and FR with transmutation scenarios respectively, and 34% for the fourth scenario. (authors)

  20. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  1. Preliminary Conceptual Design and Development of Core Technology of Very High Temperature Gas-Cooled Reactor Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jong Hwa; Kang, H. S.; Gil, C. S. and others

    2006-05-15

    For the nuclear hydrogen production system, the VHTR technology and the IS cycle technology are being developed. A comparative evaluation on the block type reactor and the pebble type reactor is performed to decide a proper nuclear hydrogen production reactor. 100MWt prismatic type reactor is tentatively decided and its safety characteristics are roughly investigated. Computation codes of nuclear design, thermo-fluid design, safety-performance analysis are developed and verified. Also, the development of a risk informed design technology is started. Experiments for metallic materials and graphites are carried out for the selection of materials of VHTR components. Diverse materials for process heat exchanger are studied in various corrosive environments. Pyrolytic carbon and SiC coating technology is developed and fuel manufacturing technology is basically established. Computer program is developed to evaluate the performance of coated particle fuels.

  2. Neutron cross-sections for advanced nuclear systems: the n_TOF project at CERN

    Directory of Open Access Journals (Sweden)

    Barbagallo M.

    2014-01-01

    Full Text Available The study of neutron-induced reactions is of high relevance in a wide variety of fields, ranging from stellar nucleosynthesis and fundamental nuclear physics to applications of nuclear technology. In nuclear energy, high accuracy neutron data are needed for the development of Generation IV fast reactors and accelerator driven systems, these last aimed specifically at nuclear waste incineration, as well as for research on innovative fuel cycles. In this context, a high luminosity Neutron Time Of Flight facility, n_TOF, is operating at CERN since more than a decade, with the aim of providing new, high accuracy and high resolution neutron cross-sections. Thanks to the features of the neutron beam, a rich experimental program relevant to nuclear technology has been carried out so far. The program will be further expanded in the near future, thanks in particular to a new high-flux experimental area, now under construction.

  3. Nuclear physics with advanced brilliant gamma beams at ELI-NP

    Science.gov (United States)

    Ur, Călin A.; Filipescu, Dan; Gheorghe, Ioana; Iancu, Violeta; Suliman, Gabriel; Teşileanu, Ovidiu

    2016-01-01

    The Extreme Light Infrastructure - Nuclear Physics facility is dedicated to nuclear physics studies with the use of extreme electromagnetic radiation. One of the main research system to be installed and operated in the facility is an outstanding high brilliance gamma beam system. The Gamma Beam System of ELI-NP will produce intense, quasi-monochromatic gamma beams via inverse Compton scattering of short laser pulses on relativistic electron beam pulses. The gamma beams available at ELI-NP will allow for the performance of photo-nuclear reactions aiming to reveal the intimate structure of the atomic nucleus. Nuclear Resonance Fluorescence, photo-fission, photo-disintegration reactions above the particle threshold will be used to study the dipole response of nuclei, the structure of the Pygmy resonances, nuclear processes relevant for astrophysics, production and study of exotic neutron-rich nuclei.

  4. Nuclear physics with advanced brilliant gamma beams at ELI–NP

    Directory of Open Access Journals (Sweden)

    Ur Călin A.

    2016-01-01

    Full Text Available The Extreme Light Infrastructure - Nuclear Physics facility is dedicated to nuclear physics studies with the use of extreme electromagnetic radiation. One of the main research system to be installed and operated in the facility is an outstanding high brilliance gamma beam system. The Gamma Beam System of ELI–NP will produce intense, quasi–monochromatic gamma beams via inverse Compton scattering of short laser pulses on relativistic electron beam pulses. The gamma beams available at ELI–NP will allow for the performance of photo-nuclear reactions aiming to reveal the intimate structure of the atomic nucleus. Nuclear Resonance Fluorescence, photo-fission, photo-disintegration reactions above the particle threshold will be used to study the dipole response of nuclei, the structure of the Pygmy resonances, nuclear processes relevant for astrophysics, production and study of exotic neutron–rich nuclei.

  5. Neutron cross-sections for advanced nuclear systems: the n_TOF project at CERN

    Science.gov (United States)

    Barbagallo, M.; Mastromarco, M.; Colonna, N.; Altstadt, S.; Andrzejewski, J.; Audouin, L.; Bécares, V.; Bečvář, F.; Belloni, F.; Berthoumieux, E.; Billowes, J.; Bosnar, D.; Brugger, M.; Calviani, M.; Calviño, F.; Cano-Ott, D.; Carrapiço, C.; Cerutti, F.; Chiaveri, E.; Chin, M.; Cortés, G.; Cortés-Giraldo, M. A.; Diakaki, M.; Domingo-Pardo, C.; Duran, I.; Dressler, R.; Eleftheriadis, C.; Ferrari, A.; Fraval, K.; Ganesan, S.; García, A. R.; Giubrone, G.; Gonçalves, I. F.; González-Romero, E.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Hernández-Prieto, A.; Jenkins, D. G.; Jericha, E.; Kadi, Y.; Käppeler, F.; Karadimos, D.; Kivel, N.; Koehler, P.; Krtička, M.; Kroll, J.; Lampoudis, C.; Langer, C.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Leong, L. S.; Losito, R.; Manousos, A.; Marganiec, J.; Martínez, T.; Massimi, C.; Mastinu, P. F.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Mondalaers, W.; Paradela, C.; Pavlik, A.; Perkowski, J.; Plompen, A.; Praena, J.; Quesada, J. M.; Rauscher, T.; Reifarth, R.; Riego, A.; Rubbia, C.; Sabaté-Gilarte, M.; Sarmento, R.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Tagliente, G.; Tain, J. L.; Tarrío, D.; Tassan-Got, L.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Vermeulen, M. J.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Ware, T.; Weigand, M.; Weiß, C.; Wright, T.; Žugec, P.

    2014-12-01

    The study of neutron-induced reactions is of high relevance in a wide variety of fields, ranging from stellar nucleosynthesis and fundamental nuclear physics to applications of nuclear technology. In nuclear energy, high accuracy neutron data are needed for the development of Generation IV fast reactors and accelerator driven systems, these last aimed specifically at nuclear waste incineration, as well as for research on innovative fuel cycles. In this context, a high luminosity Neutron Time Of Flight facility, n_TOF, is operating at CERN since more than a decade, with the aim of providing new, high accuracy and high resolution neutron cross-sections. Thanks to the features of the neutron beam, a rich experimental program relevant to nuclear technology has been carried out so far. The program will be further expanded in the near future, thanks in particular to a new high-flux experimental area, now under construction.

  6. Scoping calculations of power sources for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Difilippo, F.C. [Oak Ridge National Lab., TN (United States)

    1994-05-01

    This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to making scoping calculations for mission analysis.

  7. NUCLEAR HYDROGEN DEVERLOPMENT AND DEMONSTRATION PROGRAM IN KOREA

    Institute of Scientific and Technical Information of China (English)

    林昌奎; 朴原奭; 张锺和

    2004-01-01

    The Korea Atomic Energy Research Institute (KAERI) launched a nuclear hydrogen program to develop and demonstrate a hydrogen mass production system by 2019. The key feature of the system is to produce the hydrogen from water using nuclear power. No fossil fuel (energy) will be used and no greenhouse gas will be emitted. The high temperature gas-cooled reactor (HTGR), which has high safety characteristics with competitive economy, is being developed to provide a high temperature heat (about 1000℃) requir...

  8. Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tome, Carlos N [Los Alamos National Laboratory; Caro, J A [Los Alamos National Laboratory; Lebensohn, R A [Los Alamos National Laboratory; Unal, Cetin [Los Alamos National Laboratory; Arsenlis, A [LLNL; Marian, J [LLNL; Pasamehmetoglu, K [INL

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.

  9. The Plant-Window System: A framework for an integrated computing environment at advanced nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, R.T.; Mullens, J.A. [Oak Ridge National Lab., TN (United States); Naser, J.A. [Electric Power Research Inst., Palo Alto, CA (United States)

    1997-10-01

    Power plant data, and the information that can be derived from it, provide the link to the plant through which the operations, maintenance and engineering staff understand and manage plant performance. The extensive use of computer technology in advanced reactor designs provides the opportunity to greatly expand the capability to obtain, analyze, and present data about the plant to station personnel. However, to support highly efficient and increasingly safe operation of nuclear power plants, it is necessary to transform the vast quantity of available data into clear, concise, and coherent information that can be readily accessed and used throughout the plant. This need can be met by an integrated computer workstation environment that provides the necessary information and software applications, in a manner that can be easily understood and sued, to the proper users throughout the plan. As part of a Cooperative Research and Development Agreement with the Electric Power Research Institute, the Oak Ridge National laboratory has developed functional requirements for a Plant-Wide Integrated Environment Distributed On Workstations (Plant-Window) System. The Plant-Window System (PWS) can serve the needs of operations, engineering, and maintenance personnel at nuclear power stations by providing integrated data and software applications within a common computing environment. The PWS requirements identify functional capabilities and provide guidelines for standardized hardware, software, and display interfaces so as to define a flexible computing environment for both current generation nuclear power plants and advanced reactor designs.

  10. Nonlinear adaptive power-level control for modular high temperature gas-cooled reactors%模块式高温气冷堆的非线性自适应功率控制

    Institute of Scientific and Technical Information of China (English)

    董哲

    2013-01-01

    Because of its inherent safety feature and potential economic competitiveness,the modular high temperature gas-cooled reactor (MHTGR) is regarded as the central part of the next generation of nuclear plant(NGNP).Power-level control is one of the key techniques that guarantee the safe,stable and efficient operation for any nuclear reactors.Since the MHTGR dynamics have the characteristics of strong nonlinearity and uncertainty,which can improve the operation performance.It is significant to develop the nonlinear adaptive power-level control er for the MHTGR.Based on the natural dynamic features beneficial to system stabilization,a novel nonlinear adaptive power-level controller is given for the MHTGR in this paper.It has proved theoretical y that this newly-built control er not only provides the overal asymptotic closed-loop stability,but also is adaptive to the system uncertainty.This control er has been applied in the power-level regulation of the pebble-bed MHTGR of the HTR-PM power plant.Numerical simulation results confirm the feasibility of this control strategy and the relationship between the performance and parameters.%由于具有固有安全性和潜在经济竞争力,模块式高温气冷堆(Modular High Temperature Gas-cooled Reactor, MHTGR)已被视为下一代核能系统的首选堆型之一。功率调节是保证MHTGR安全、稳定和高效运行的关键技术之一,而MHTGR在动态特性上具有非线性强和不确定性大的特点,因此发展MHTGR的非线性自适应功率控制器对于提升运行性能具有重要意义。在充分利用系统自身对镇定有益的动态特性的基础上,本文给出一种新颖的MHTGR非线性自适应功率控制器,并从理论上证明该控制器既可以保证闭环系统的全局渐近稳定性,又对系统不确定性具有自适应功能。数值仿真结果不仅验证了控制器的可行性,而且还给出了调节性能与控制器参数间的关系。

  11. 75 FR 38151 - Governors' Designees Receiving Advance Notification of Transportation of Nuclear Waste

    Science.gov (United States)

    2010-07-01

    ... to Dr. Stephen N. Salomon, Office of Federal and State Materials and Environmental Management Programs, U.S. Nuclear Regulatory Commission, Washington, DC 20555, by e-mail at Stephen.Salomon@nrc.gov...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-12-01

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

  13. World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

    Energy Technology Data Exchange (ETDEWEB)

    1979-06-01

    Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

  14. Advanced Quantification of Plutonium Ionization Potential to Support Nuclear Forensic Evaluations by Resonance Ionization Mass Spectrometry

    Science.gov (United States)

    2015-06-01

    decay and induced emission and absorption are almost identical. Their probability distributions are closely related. The electromagnetic radiation ...NAVAL POSTGRADUATE SCHOOL MONTEREY, CALIFORNIA THESIS Approved for public release; distribution is unlimited ADVANCED ...ONLY (Leave blank) 2. REPORT DATE June 2015 3. REPORT TYPE AND DATES COVERED Master’s Thesis 4. TITLE AND SUBTITLE ADVANCED QUANTIFICATION OF

  15. Preliminary Study of Gas Cooled Fast Breeder Reactor with Heterogen Percentage of Uranium–Plutonium Carbide based fuel and 300 MWt Power

    Science.gov (United States)

    Clief Pattipawaej, Sandro; Su’ud, Zaki

    2017-01-01

    A preliminary design study of GFR with helium gas-cooled has been performed. In this study used natural uranium and plutonium results LWR waste as fuel. Fuel with a small percentage of plutonium are arranged on the inside of the core area, and the fuel with a greater percentage set on the outside of the core area. The configuration of such fuel is deliberately set to increase breeding in this part of the central core and reduce the leakage of neutrons on the outer side of the core, in order to get long-lived reactor with a small reactivity. Configuration of fuel as it is also useful to generate a peak power reactors with relatively low in both the direction of axial or radial. Optimization has been done to fuel fraction 45.0% was found that the reactor may be operating in more than 10 year time with excess reactivity less than 1%.

  16. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  17. Analytical Solution of Fick's Law of the TRISO-Coated Fuel Particles and Fuel Elements in Pebble-Bed High Temperature Gas-Cooled Reactors

    Institute of Scientific and Technical Information of China (English)

    CAO Jian-Zhu; FANG Chao; SUN Li-Feng

    2011-01-01

    T wo kinds of approaches are built to solve the fission products diffusion models (Fick's equation) based on sphere fuel particles and sphere fuel elements exactly. Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented, respectively. The analytica,solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system, a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element. Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.%@@ Two kinds of approaches are built to solve the fission products diffusion models(Fick's equation) based on sphere fuel particles and sphere fuel elements exactly.Two models for homogenous TRISO-coated fuel particles and fuel elements used in pebble-bed high temperature gas-cooled reactors are presented,respectively.The analytical solution of Fick's equation for fission products diffusion in fuel particles is derived by variables separation.In the fuel element system,a modification of the diffusion coefficient from D to D/r is made to characterize the difference of diffusion rates in distinct areas and it is shown that the Laplace and Hankel transformations are effective as the diffusion coefficient in Fick's equation is dependant on the radius of the fuel element.Both the solutions are useful for the prediction of the fission product behaviors and could be programmed in the corresponding engineering calculations.

  18. FOREWORD: International Summer School for Advanced Studies 'Dynamics of open nuclear systems' (PREDEAL12)

    Science.gov (United States)

    Delion, D. S.; Zamfir, N. V.; Raduta, A. R.; Gulminelli, F.

    2013-02-01

    This proceedings volume contains the invited lectures and contributions presented at the International Summer School on Nuclear Physics held at Trei Brazi, a summer resort of the Bioterra University, near the city of Predeal, Romania, on 9-20 July 2012. The long tradition of International Summer Schools on Nuclear Physics in Romania dates as far back as 1964, with the event being scheduled every two years. During this period of almost 50 years, many outstanding nuclear scientists have lectured on various topics related to nuclear physics and particle physics. This year we celebrate the 80th birthday of Aureliu Sandulescu, one of the founders of the Romanian school of theoretical nuclear physics. He was Serban Titeica's PhD student, one of Werner Heisenberg's PhD students, and he organized the first edition of this event. Aureliu Sandulescu's major contributions to the field of theoretical nuclear physics are related in particular to the prediction of cluster radioactivity, the physics of open quantum systems and the innovative technique of detecting superheavy nuclei using the double magic projectile 48Ca (Calcium), nowadays a widely used method at the JINR—Dubna and GSI—Darmstadt laboratories. The title of the event, 'Dynamics of Open Nuclear Systems', is in recognition of Aureliu Sandulescu's great personality. The lectures were attended by Romanian and foreign Master and PhD students and young researchers in nuclear physics. About 25 reputable professors and researchers in nuclear physics delivered lectures during this period. According to a well-established tradition, an interval of two hours was allotted for each lecture (including discussions). Therefore we kept a balance between the school and conference format. Two lectures were held during the morning and afternoon sessions. After lecture sessions, three or four oral contributions were given by young scientists. This was a good opportunity for them to present the results of their research in front of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

  20. Ship-Based Nuclear Energy Systems for Accelerating Developing World Socioeconomic Advance

    Science.gov (United States)

    Petroski, Robert; Wood, Lowell

    2014-07-01

    Technological, economic, and policy aspects of supplying energy to newly industrializing and developing countries using ship-deployed nuclear energy systems are described. The approach analyzed comprises nuclear installations of up to gigawatt scale deployed within currently mass-produced large ship hulls which are capable of flexibly supplying energy for electricity, water desalination and district heating-&-cooling with low latencies and minimized shoreside capital expenditures. Nuclear energy is uniquely suited for mobile deployment due to its combination of extraordinary energy density and high power density, which enable enormous supplies of energy to be deployed at extremely low marginal costs. Nuclear installations on ships also confer technological advantages by essentially eliminating risk from earthquakes, tsunamis, and floods; taking advantage of assured access to an effectively unlimited amount of cooling water, and involving minimal onshore preparations and commitments. Instances of floating nuclear power stations that have been proposed in the past, some of which are currently being pursued, have generally been based on conventional LWR technology, moreover without flexibility or completeness of power output options. We consider nuclear technology options for their applicability to the unique opportunities and challenges of a marine environment, with special attention given to low-pressure, high thermal margin systems with continuous and assured afterheat dissipation into the ambient seawater. Such systems appear promising for offering an exceptionally high degree of safety while using a maximally simple set of components. We furthermore consider systems tailored to Developing World contexts, which satisfy societal requirements beyond electrification, e.g., flexible sourcing of potable water and HVAC services, servicing time-varying user requirements, and compatibility with the full spectrum of local renewable energy supplies, specifically including

  1. THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FOR VARIOUS PROLIFERATION AND THEFT SCENARIOS

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Ebbinghaus, Bartley B.; Collins, Brian A.; Sleaford, Brad W.; Hase, Kevin R.; Robel, Martin; Wallace, R. K.; Bradley, Keith S.; Ireland, J. R.; Jarvinen, G. D.; Johnson, M. W.; Prichard, Andrew W.; Smith, Brian W.

    2012-08-29

    We must anticipate that the day is approaching when details of nuclear weapons design and fabrication will become common knowledge. On that day we must be particularly certain that all special nuclear materials (SNM) are adequately accounted for and protected and that we have a clear understanding of the utility of nuclear materials to potential adversaries. To this end, this paper examines the attractiveness of materials mixtures containing SNM and alternate nuclear materials associated with the plutonium-uranium reduction extraction (Purex), uranium extraction (UREX), coextraction (COEX), thorium extraction (THOREX), and PYROX (an electrochemical refining method) reprocessing schemes. This paper provides a set of figures of merit for evaluating material attractiveness that covers a broad range of proliferant state and subnational group capabilities. The primary conclusion of this paper is that all fissile material must be rigorously safeguarded to detect diversion by a state and must be provided the highest levels of physical protection to prevent theft by subnational groups; no 'silver bullet' fuel cycle has been found that will permit the relaxation of current international safeguards or national physical security protection levels. The work reported herein has been performed at the request of the U.S. Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for, the nuclear materials in DOE nuclear facilities. The methodology and findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security is discussed.

  2. The attractiveness of materials in advanced nuclear fuel cycles for various proliferation and theft scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, Charles G [Los Alamos National Laboratory; Wallace, Richard K [Los Alamos National Laboratory; Ireland, John R [Los Alamos National Laboratory; Johnson, M W [Los Alamos National Laboratory; Hase, Kevin R [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Ebbinghaus, Bartley B [LLNL; Sleaford, Brad A [LLNL; Bradley, Keith S [LLNL; Collins, Brian W [PNNL; Smith, Brian W [PNNL; Prichard, Andrew W [PNNL

    2009-01-01

    This paper is an extension to earlier studies that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, COEX, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant state and sub-national group capabilities. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a state and provided the highest levels of physical protection to prevent theft by sub-national groups; no 'silver bullet' has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities. The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  3. U.S. Department of Energy facilities needed to advance nuclear power.

    Science.gov (United States)

    Ahearne, John F

    2011-01-01

    This talk is based upon a November 2008 report by the U.S. Department of Energy (DOE) Nuclear Energy Advisory Committee (NEAC). The report has two parts, a policy section and a technology section. Here extensive material from the Technical Subcommittee section of the NEAC report is used.

  4. Development of Advanced Technologies to Reduce Design, Fabrication and Construction Costs for Future Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    DiNunzio, Camillo A. [Framatome ANP DE& S, Marlborough, MA (United States); Gupta, Abhinav [Univ. of North Carolina, Raleigh, NC (United States); Golay, Michael [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Luk, Vincent [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Turk, Rich [Westinghouse Electric Company Nuclear Systems, Windsor, CT (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jin, Geum-Taek [Korea Power Engineering Company Inc., Yongin-si, Gyeonggi-do (Korea, Republic of)

    2002-11-30

    This report presents a summation of the third and final year of a three-year investigation into methods and technologies for substantially reducing the capital costs and total schedule for future nuclear plants. In addition, this is the final technical report for the three-year period of studies.

  5. FURTHER ASSESSMENTS OF THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FROM A SAFEGUARDS PERSPECTIVE

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Jarvinen, G. D.; Wallace, R. K.; Ireland, J. R.; Johnson, M. W.; Sleaford, Brad W.; Ebbinghaus, B. B.; Bradley, Keith S.; Collins, Brian A.; Smith, Brian W.; Prichard, Andrew W.

    2008-10-01

    This paper summarizes the results of an extension to an earlier study [ ] that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with the PUREX, UREX+, and COEX reprocessing schemes. This study focuses on the materials associated with the UREX, COEX, THOREX, and PYROX reprocessing schemes. This study also examines what is required to render plutonium as “unattractive.” Furthermore, combining the results of this study with those from the earlier study permits a comparison of the uranium and thorium based fuel cycles on the basis of the attractiveness of the SNM associated with each fuel cycle. Both studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of “attractiveness levels” that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities [ ]. The methodology and key findings will be presented. Additionally, how these attractiveness levels relate to proliferation resistance (e.g. by increasing impediments to the diversion, theft, undeclared production of SNM for the purpose of acquiring a nuclear weapon), and how they could be used to help inform policy makers, will be discussed.

  6. The Attractiveness of Materials in Advanced Nuclear Fuel Cycles for Various Proliferation and Theft Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Bathke, C. G.; Wallace, R. K.; Ireland, J. R.; Johnson, M. W.; Hase, Kevin R.; Jarvinen, G. D.; Ebbinghaus, B. B.; Sleaford, Brad W.; Bradley, Keith S.; Collins, Brian A.; Smith, Brian W.; Prichard, Andrew W.

    2010-09-01

    This paper is an extension to earlier studies1,2 that examined the attractiveness of materials mixtures containing special nuclear materials (SNM) and alternate nuclear materials (ANM) associated with the PUREX, UREX, COEX, THOREX, and PYROX reprocessing schemes. This study extends the figure of merit (FOM) for evaluating attractiveness to cover a broad range of proliferant state and sub-national group capabilities. The primary conclusion of this study is that all fissile material needs to be rigorously safeguarded to detect diversion by a state and provided the highest levels of physical protection to prevent theft by sub-national groups; no “silver bullet” has been found that will permit the relaxation of current international safeguards or national physical security protection levels. This series of studies has been performed at the request of the United States Department of Energy (DOE) and is based on the calculation of "attractiveness levels" that are expressed in terms consistent with, but normally reserved for nuclear materials in DOE nuclear facilities.3 The expanded methodology and updated findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security are discussed.

  7. An Investigation for Arranging the Video Display Unit Information in a Main Control Room of Advanced Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Chong Cheng; Yang, Chih Wei [Institute of Nuclear Energy Research, Atomic Energy Council, Taoyuan (China)

    2014-08-15

    Current digital instrumentation and control and main control room (MCR) technology has extended the capability of integrating information from numerous plant systems and transmitting needed information to operations personnel in a timely manner that could not be envisioned when previous generation plants were designed and built. A MCR operator can complete all necessary operating actions on the video display unit (VDU). It is extremely flexible and convenient for operators to select and to control the system display on the screen. However, a high degree of digitalization has some risks. For example, in nuclear power plants, failures in the instrumentation and control devices could stop the operation of the plant. Human factors engineering (HFE) approaches would be a manner to solve this problem. Under HFE considerations, there exists 'population stereotype' for operation. That is, the operator is used to operating a specific display on the specific VDU for operation. Under emergency conditions, there is possibility that the operator will response with this habit population stereotype, and not be aware that the current situation has already changed. Accordingly, the advanced nuclear power plant should establish the MCR VDU configuration plan to meet the consistent teamwork goal under normal operation, transient and accident conditions. On the other hand, the advanced nuclear power plant should establish the human factors verification and validation plan of the MCR VDU configuration to verify and validate the configuration of the MCR VDUs, and to ensure that the MCR VDU configuration allows the operator shift to meet the HFE consideration and the consistent teamwork goal under normal operation, transient and accident conditions. This paper is one of the HF V V plans of the MCR VDU configuration of the advanced nuclear power plant. The purpose of this study is to confirm whether the VDU configuration meets the human factors principles and the consistent

  8. A study on advanced man-machine interface system for autonomous nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Matsuoka, Takeshi; Numano, Masayoshi; Someya, Minoru; Fukuto, Junji; Mitomo, Noboru; Miyazaki, Keiko; Sugasawa, Shinobu [Ship Research Inst., Mitaka, Tokyo (Japan)

    1997-01-01

    Research on Artificial Intelligence Systems for Nuclear Installations has been performed in cooperation with five research institutes (Ship Research Institute, Electrotechnical Laboratory, Japan Atomic Energy Research Institute, Power Reactor and Nuclear Fuel Development Corporation, The Institute of Physical and Chemical Research), from 1989 to 1994 as the Cross-over Research Group with the support of the Science and Technology Agency. Ship Research Institute has been carrying out the research on the Man-Machine Interface (MMI) system for autonomous nuclear power plants. This paper describes the concept of autonomous nuclear power plants, a plant simulator of an autonomous nuclear power plant, a contracted function model of a plant state, three-dimensional color graphic display of a plant state, and a function of automatic classification of plant states by the COBWEB method. A plant simulator has been developed by using the expert system G2 (Gensym Co.). The simulator generates plant process data at each component of a plant. This simulator models a pressurized water reactor and some examples of autonomous functions are incorporated. A contracted function model of a plant state has been produced at the main part of the MMI system based on plant process data from the simulator. The main purpose of the present study is to give the MMI system a function to identify the plant operational state, to update and revise the function model, and to expand a knowledge. A plant state is expressed in a three-dimensional graphic display which receives sensor values from the plant simulator and expresses the plant state in nearly real time speed. A research on the automatic classification of plant states has been also performed, which shows us the relations among different plant states. The study is being continued to the 2nd stage Cross-over Research from 1994, as the Study on Divers, Cooperative Intelligent System for Autonomous Plants. (J.P.N.)

  9. Advanced x-ray spectrometric techniques for characterization of nuclear materials: An overview of recent laboratory activities

    Science.gov (United States)

    Misra, N. L.

    2014-11-01

    Advancements in x-ray spectrometric techniques at different stages have made this technique suitable for characterization of nuclear materials with respect to trace/major element determinations and compositional uniformity studies. The two important features of total reflection x-ray fluorescence spectrometry: 1) requirement of very small amount of sample in ng level 2) multielement analytical capability, in addition to other features, make this technique very much suitable to nuclear materials characterization as most of the nuclear materials are radioactive and the radioactive waste generated and radiation hazards to the operator are minimum when such low amount of sample is used. Similarly advanced features of energy dispersive x-ray fluorescence e.g. better geometry for high flux, reduction in background due to application of radiation filters have made the measurements of samples sealed inside thin alkathene/PVC covers possible with good sensitivity. This approach avoids putting the instrument inside a glove box for measuring radioactive samples and makes the operation/maintenance of the instrument and analysis of the samples possible in easy and fast manner. This approach has been used for major element determinations in mixed uranium-plutonium samples. Similarly μ-XRF with brilliant and micro-focused excitation sources can be used for compositional uniformity study of reactor fuel pellets. A μ-XRF study using synchrotron light source has been made to assess the compositional uniformity of mixed uranium-thorium oxide pellets produced by different processes. This approach is simple as it does not involve any sample preparation and is non-destructive. A brief summary of such activities carried out in our laboratory in past as well as ongoing and planned for the future have been discussed in the present manuscript.

  10. Study of Physical Scheme for 10 MW Accelerator-driven Fast-thermal Coupled Gas-cooled Reactor%加速器驱动10 MW快热耦合气冷堆物理方案研究

    Institute of Scientific and Technical Information of China (English)

    李金阳; 顾龙; 秦长平; 王大伟; 刘璐

    2013-01-01

    The accelerator-driven sub-critical system has promising future in transmuta-tion of nuclear spent fuels .A physical design of a 10MW fast-thermal spectrum gas-cooled reactor was studied .The program COUPLE2.0 ,which couples with MCNPX and ORIGEN ,was applied to compute this scheme ,and the continuous energy neutron cross section was obtained by the nuclear data library of ENDF-7 which was modified with five different temperatures :300 ,600 ,900 ,1 500 ,and 2500K .The physical pa-rameters such as kef ,proton efficiency ,delayed neutron fraction and accelerator current varying with the burnup time were obtained .The system during 350 d burnup was obtained and the transmutation capability was analyzed by the further calculation .It is found that during the 350 d burnup ,the variations of kef and accelerator current are relatively small .The system has good transmutation capability with the transmutation support ratio of 24.86 .%加速器驱动的次临界系统(ADS )在实现嬗变核废料方面具有良好的前景。对加速器驱动10 M W次临界快热耦合能谱气冷堆的物理方案进行了设计和研究,利用 MCNPX和ORIGEN耦合的计算程序COUPLE2.0对该方案进行了计算,其中,中子截面采用ENDF-7处理后得到的5个温度300、600、900、1500、2500 K下的连续能量核数据库。得到该方案350 d燃耗期间的 kef 、质子效率、缓发中子份额以及加速器束流强度的变化。进而计算得到了该方案的燃耗信息,并分析了该方案的嬗变能力。结果表明,该方案在350 d燃耗期间的 kef 、加速器束流强度变化较小,嬗变支持比为24.86,具有较好的嬗变效果。

  11. 高温气冷堆堆芯实时热工水力模型%Real Time Thermal Hydraulic Model for High Temperature Gas-cooled Reactor Core

    Institute of Scientific and Technical Information of China (English)

    眭喆; 张瑞鹏; 孙俊; 马远乐

    2013-01-01

    A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model.%为建立适用于球床式高温气冷堆核电厂的模拟机,采用一体化仿真支撑平台vPower建立高温气冷堆堆芯的实时热工水力模型,利用流体网络求解氦气流道的流量与压力分布及传热网络求解球床燃料区、石墨反射层区与碳砖区的温度分布,实现整个氦气流场与固相温度场的实时、耦合计算.模拟100%额定负荷和50%额定负荷2个稳态工况和入口温度阶跃和流量阶跃2个动态过程.稳态工况与设计参数的定量对比以及动态过程的定性分析表明,该模型具有较好的适用性.

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

    Energy Technology Data Exchange (ETDEWEB)

    G. Palmiotti; M. Salvatores; H. Hiruta

    2011-06-01

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

  13. The development and use of radionuclide generators in nuclear medicine -- recent advances and future perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Knapp, F.F. Jr.

    1998-03-01

    Although the trend in radionuclide generator research has declined, radionuclide generator systems continue to play an important role in nuclear medicine. Technetium-99m obtained from the molybdenum-99/technetium-99m generator system is used in over 80% of all diagnostic clinical studies and there is increasing interest and use of therapeutic radioisotopes obtained from generator systems. This paper focuses on a discussion of the major current areas of radionuclide generator research, and the expected areas of future research and applications.

  14. Recent advances and results from the solid radiochemistry nuclear diagnostic at the National Ignition Facility

    Science.gov (United States)

    Gharibyan, N.; Shaughnessy, D. A.; Moody, K. J.; Grant, P. M.; Despotopulos, J. D.; Faye, S. A.; Jedlovec, D. R.; Yeamans, C. B.

    2016-11-01

    The solid debris collection capability at the National Ignition Facility has been expanded to include a third line-of-sight assembly. The solid radiochemistry nuclear diagnostic measurement of the ratio of gold isotopes is dependent on the efficient collection of neutron-activated hohlraum debris by passive metal disks. The collection of target debris at this new location is more reliable in comparison to the historic locations, and it appears to be independent of collector surface ablation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-04-01

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

  16. Requirements for a Dynamic Solvent Extraction Module to Support Development of Advanced Technologies for the Recycle of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jack Law; Veronica Rutledge; Candido Pereira; Jackie Copple; Kurt Frey; John Krebs; Laura Maggos; Kevin Nichols; Kent Wardle; Pratap Sadasivan; Valmor DeAlmieda; David Depaoli

    2011-06-01

    The Department of Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has been established to create and deploy next generation, verified and validated nuclear energy modeling and simulation capabilities for the design, implementation, and operation of future nuclear energy systems to improve the U.S. energy security. As part of the NEAMS program, Integrated Performance and Safety Codes (IPSC's) are being produced to significantly advance the status of modeling and simulation of energy systems beyond what is currently available to the extent that the new codes be readily functional in the short term and extensible in the longer term. The four IPSC areas include Safeguards and Separations, Reactors, Fuels, and Waste Forms. As part of the Safeguards and Separations (SafeSeps) IPSC effort, interoperable process models are being developed that enable dynamic simulation of an advanced separations plant. A SafeSepss IPSC 'toolkit' is in development to enable the integration of separation process modules and safeguards tools into the design process by providing an environment to compose, verify and validate a simulation application to be used for analysis of various plant configurations and operating conditions. The modules of this toolkit will be implemented on a modern, expandable architecture with the flexibility to explore and evaluate a wide range of process options while preserving their stand-alone usability. Modules implemented at the plant-level will initially incorporate relatively simple representations for each process through a reduced modeling approach. Final versions will incorporate the capability to bridge to subscale models to provide required fidelity in chemical and physical processes. A dynamic solvent extraction model and its module implementation are needed to support the development of this integrated plant model. As a stand-alone application, it will also support solvent development of extraction flowsheets

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

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1993-08-11

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

  18. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2011-10-01

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

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

    Science.gov (United States)

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

    1975-01-01

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

  20. Idaho National Laboratory Experimental Program to Measure the Flow Phenomena in a Scaled Model of a Prismatic Gas-Cooled Reactor Lower Plenum for Validation of CFD Codes

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

    2008-09-01

    The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow - with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements

  1. Observations of asymmetric velocity fields and gas cooling in the NGC 4636 galaxy group X-ray halo

    CERN Document Server

    Ahoranta, Jussi; Pinto, Ciro; Sanders, Jeremy; Kaastra, Jelle; de Plaa, Jelle; Fabian, Andrew

    2016-01-01

    This study aims to probe the thermodynamic properties of the hot intragroup medium (IGM) plasma in the core regions of the NGC 4636 galaxy group by detailed measurements of several emission lines and their relative intensities. We analyzed deep XMM-Newton Reflection Grating Spectrometer (RGS) data in five adjacent spectral regions in the central parts of the NGC 4636 galaxy group. We examined the suppression of the Fe xvii resonance line (15.01 {\\AA}) as compared to the forbidden lines of the same ion (17.05 {\\AA} and 17.10 {\\AA}). The presence and radial dependence of the cooling flow was investigated through spectral modeling. In addition, a parallel analysis with deep Chandra Advances CCD Imaging Spectrometer (ACIS) data was conducted to gain additional information about the thermodynamical properties of the IGM. We find that the plasma at the group center to the north shows efficient Fe xvii ion resonant scattering, wheras no resonant scattering was detected at the south side. The regions featuring resona...

  2. OSMOSE an experimental program for improving neutronic predictions of advanced nuclear fuels.

    Energy Technology Data Exchange (ETDEWEB)

    Klann, R. T.; Aliberti, G.; Zhong, Z.; Graczyk, D.; Loussi, A.; Nuclear Engineering Division; Commissariat a l Energie Atomique

    2007-10-18

    This report describes the technical results of tasks and activities conducted in FY07 to support the DOE-CEA collaboration on the OSMOSE program. The activities are divided into five high-level tasks: reactor modeling and pre-experiment analysis, sample fabrication and analysis, reactor experiments, data treatment and analysis, and assessment for relevance to high priority advanced reactor programs (such as GNEP and Gen-IV).

  3. Multiscale Modeling of the Deformation of Advanced Ferritic Steels for Generation IV Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Nasr M. Ghoniem; Nick Kioussis

    2009-04-18

    The objective of this project is to use the multi-scale modeling of materials (MMM) approach to develop an improved understanding of the effects of neutron irradiation on the mechanical properties of high-temperature structural materials that are being developed or proposed for Gen IV applications. In particular, the research focuses on advanced ferritic/ martensitic steels to enable operation up to 650-700°C, compared to the current 550°C limit on high-temperature steels.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-31

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

  5. Introduction of Deterministic OS for SPLC in Advanced Nuclear I and C System

    Energy Technology Data Exchange (ETDEWEB)

    Son, C. W.; Kim, D. H.; Son, G. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Existing PLC's either had no output logic with regard to devices' redundant structure or it was set in a fixed way, and as a result it was extremely inefficient to use them for redundant systems such as that of a nuclear power plant and their use was limited. This project has developed a real-time operating system with Redundancy handling logic in order to improve the problems mentioned. That is the scheduling method, redundancy handling function of supervisory. A controller used in Nuclear Power Plants safety system has been developed differently from industrial PLC by its design, production and test. Therefore, the operation system that controls the NPPs safety PLC should be developed properly so that it can work well on the handling logic of the safety system. Scoops has secured the deterministic for the safety PLC control by possessing the sequential scheduling, time management, redundancy handling logic and system diagnosis function considering the connectivity of all the software module by Supervisory.

  6. Advanced biological treatment of aqueous effluent from the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Pitt, Jr., W. W.; Hancher, C. W.; Patton, B. D.; Shumate, II, S. E.

    1979-01-01

    Many of the processing steps in the nuclear fuel cycle generate aqueous effluent streams bearing contaminants that can, because of their chemical or radiological properties, pose an environmental hazard. Concentration of such contaminants must be reduced to acceptable levels before the streams can be discharged to the environment. Two classes of contaminants, nitrates and heavy metals, are addressed in this study. Specific techniques aimed at the removal of nitrates and radioactive heavy metals by biological processes are being developed, tested, and demonstrated. Although cost comparisons between biological processes and current treatment methods are presented, these comparisons may be misleading because biological processes yield environmentally better end results which are difficult to price. However, a strong case is made for the use of biological processes for removing nitrates and heavy metals fron nuclear fuel cycle effluents. The estimated costs for these methods are as low as, or lower than, those for alternate processes. In addition, the resulting disposal products - nitrogen gas, CO/sub 2/, and heavy metals incorporated into microorganisms - are much more ecologically desirable than the end products of other waste treatment methods.

  7. Taking CAIRE. 20 years of experience with the advanced emergency guidance system CAIRE for nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Brenk, Hans-Dietmar [Brenk Systemplanung GmbH, Aachen (Germany)

    2011-08-15

    It has now been almost 20 years that we, at Brenk Systemplanung GmbH (BS), triggered by the accidents at ''Three Mile Island'' in 1979 and at ''Chernobyl'' in 1986, have developed the emergency decision aid system CAIRE (Computer Aided Response to Emergencies) to assist decision makers to stand the high pressure of responsibly managing and mitigating a nuclear accident. The system is designed to work automatically on the basis of telemetric monitoring networks and started operation for the first time at 4 port sites of the French navy in 1992. In cases of incidents or accidents, it has to answer the following questions automatically: Where did the incident/accident occur? Which radionuclides are being released? What are the release rates? What is the actual radiation exposure? How will the incident/accident develop? Due to the Fukushima accident, these functionalities have reached sad relevance. Based on the demand of our client, a large NPP of a German power utility company, we are planning a modernised CAIRE system to be installed at this NPP and the Nuclear Emergency Service (KHG). Currently, the conceptional work is under way. (orig.)

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

    CERN Document Server

    Calviani, Marco; Montagnoli, G; Mastinu, P

    2009-01-01

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

  9. Advancing Inverse Sensitivity/Uncertainty Methods for Nuclear Fuel Cycle Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arbanas, G., E-mail: arbanasg@ornl.gov [Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6171 (United States); Williams, M.L.; Leal, L.C.; Dunn, M.E. [Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6171 (United States); Khuwaileh, B.A.; Wang, C.; Abdel-Khalik, H. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695-7909 (United States)

    2015-01-15

    The inverse sensitivity/uncertainty quantification (IS/UQ) method has recently been implemented in the Inverse Sensitivity/UnceRtainty Estimator (INSURE) module of the AMPX cross section processing system [M.E. Dunn and N.M. Greene, “AMPX-2000: A Cross-Section Processing System for Generating Nuclear Data for Criticality Safety Applications,” Trans. Am. Nucl. Soc. 86, 118–119 (2002)]. The IS/UQ method aims to quantify and prioritize the cross section measurements along with uncertainties needed to yield a given nuclear application(s) target response uncertainty, and doing this at a minimum cost. Since in some cases the extant uncertainties of the differential cross section data are already near the limits of the present-day state-of-the-art measurements, requiring significantly smaller uncertainties may be unrealistic. Therefore, we have incorporated integral benchmark experiments (IBEs) data into the IS/UQ method using the generalized linear least-squares method, and have implemented it in the INSURE module. We show how the IS/UQ method could be applied to systematic and statistical uncertainties in a self-consistent way and how it could be used to optimize uncertainties of IBEs and differential cross section data simultaneously. We itemize contributions to the cost of differential data measurements needed to define a realistic cost function.

  10. Ion acceleration and D-D nuclear fusion in laser-generated plasma from advanced deuterated polyethylene.

    Science.gov (United States)

    Torrisi, Lorenzo

    2014-10-23

    Deuterated polyethylene targets have been irradiated by means of a 1016 W/cm2 laser using 600 J pulse energy, 1315 nm wavelength, 300 ps pulse duration and 70 micron spot diameter. The plasma parameters were measured using on-line diagnostics based on ion collectors, SiC detectors and plastic scintillators, all employed in time-of-flight configuration. In addition, a Thomson parabola spectrometer, an X-ray streak camera, and calibrated neutron dosimeter bubble detectors were employed. Characteristic protons and neutrons at maximum energies of 3.0 MeV and 2.45 MeV, respectively, were detected, confirming that energy spectra of reaction products coming from deuterium-deuterium nuclear fusion occur. In thick advanced targets a fusion rate of the order of 2 × 108 fusions per laser shot was calculated.

  11. Ion Acceleration and D-D Nuclear Fusion in Laser-Generated Plasma from Advanced Deuterated Polyethylene

    Directory of Open Access Journals (Sweden)

    Lorenzo Torrisi

    2014-10-01

    Full Text Available Deuterated polyethylene targets have been irradiated by means of a 1016 W/cm2 laser using 600 J pulse energy, 1315 nm wavelength, 300 ps pulse duration and 70 micron spot diameter. The plasma parameters were measured using on-line diagnostics based on ion collectors, SiC detectors and plastic scintillators, all employed in time-of-flight configuration. In addition, a Thomson parabola spectrometer, an X-ray streak camera, and calibrated neutron dosimeter bubble detectors were employed. Characteristic protons and neutrons at maximum energies of 3.0 MeV and 2.45 MeV, respectively, were detected, confirming that energy spectra of reaction products coming from deuterium-deuterium nuclear fusion occur. In thick advanced targets a fusion rate of the order of 2 × 108 fusions per laser shot was calculated.

  12. Thermohydraulic Design Analysis Modeling for Korea Advanced NUclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Space exploration is a realistic and profitable goal for long-term humanity survival, although the harsh space environment imposes lots of severe challenges to space pioneers. To date, almost all space programs have relied upon Chemical Rockets (CRs) rating superior thrust level to transit from the Earth's surface to its orbit. However, CRs inherently have insurmountable barrier to carry out deep space missions beyond Earth's orbit due to its low propellant efficiency, and ensuing enormous propellant requirement and launch costs. Meanwhile, nuclear rockets typically offer at least two times the propellant efficiency of a CR and thus notably reduce the propellant demand. Particularly, a Nuclear Thermal Rocket (NTR) is a leading candidate for near-term manned missions to Mars and beyond because it satisfies a relatively high thrust as well as a high efficiency. The superior efficiency of NTRs is due to both high energy density of nuclear fuel and the low molecular weight propellant of Hydrogen (H{sub 2}) over the chemical reaction by-products. A NTR uses thermal energy released from a nuclear fission reactor to heat the H{sub 2} propellant and then exhausted the highly heated propellant through a propelling nozzle to produce thrust. A propellant efficiency parameter of rocket engines is specific impulse (I{sub s}p) which represents the ratio of the thrust over the propellant consumption rate. If the average exhaust H{sub 2} temperature of a NTR is around 3,000 K, the I{sub s}p can be achieved as high as 1,000 s as compared with only 450 - 500 s of the best CRs. For this reason, NTRs are favored for various space applications such as orbital tugs, lunar transports, and manned missions to Mars and beyond. The best known NTR development effort was conducted from 1955 to1974 under the ROVER and NERVA programs in the USA. These programs had successfully designed and tested many different reactors and engines. After these projects, the researches on NERVA derived

  13. Advanced sample environments for in situ neutron diffraction studies of nuclear materials

    Science.gov (United States)

    Reiche, Helmut Matthias

    Generation IV nuclear reactor concepts, such as the supercritical-water-cooled nuclear reactor (SCWR), are actively researched internationally. Operating conditions above the critical point of water (374°C, 22.1 MPa) and fuel core temperature that potentially exceed 1850°C put a high demand on the surrounding materials. For their safe application, it is essential to characterize and understand the material properties on an atomic scale such as crystal structure and grain orientation (texture) changes as a function of temperature and stress. This permits the refinement of models predicting the macroscopic behavior of the material. Neutron diffraction is a powerful tool in characterizing such crystallographic properties due to their deep penetration depth into condensed matter. This leads to the ability to study bulk material properties, as opposed to surface effects, and allows for complex sample environments to study e.g. the individual contributions of thermo-mechanical processing steps during manufacturing, operating or accident scenarios. I present three sample environments for in situ neutron diffraction studies that provide such crystallographic information and have been successfully commissioned and integrated into the user program of the High Pressure -- Preferred Orientation (HIPPO) diffractometer at the Los Alamos Neutron Science Center (LANSCE) user facility. I adapted a sample changer for reliable and fast automated texture measurements of multiple specimens. I built a creep furnace combining a 2700 N load frame with a resistive vanadium furnace, capable of temperatures up to 1000°C, and manipulated by a pair of synchronized rotation stages. This combination allows following deformation and temperature dependent texture and strain evolutions in situ. Utilizing the presented sample changer and creep furnace we studied pressure tubes made of Zr-2.5wt%Nb currently employed in CANDURTM nuclear reactors and proposed for future SCWRs, acting as the primary

  14. A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

    Science.gov (United States)

    Anghaie, S.; Chen, G.

    1996-01-01

    A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high

  15. Inherently safe reactors and a second nuclear era.

    Science.gov (United States)

    Weinberg, A M; Spiewak, I

    1984-06-29

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

  16. Advanced Outage and Control Center: Strategies for Nuclear Plant Outage Work Status Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Gregory Weatherby

    2012-05-01

    The research effort is a part of the Light Water Reactor Sustainability (LWRS) Program. LWRS is a research and development program sponsored by the Department of Energy, performed in close collaboration with industry to provide the technical foundations for licensing and managing the long-term, safe and economical operation of current nuclear power plants. The LWRS Program serves to help the US nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. The Outage Control Center (OCC) Pilot Project was directed at carrying out the applied research for development and pilot of technology designed to enhance safe outage and maintenance operations, improve human performance and reliability, increase overall operational efficiency, and improve plant status control. Plant outage management is a high priority concern for the nuclear industry from cost and safety perspectives. Unfortunately, many of the underlying technologies supporting outage control are the same as those used in the 1980’s. They depend heavily upon large teams of staff, multiple work and coordination locations, and manual administrative actions that require large amounts of paper. Previous work in human reliability analysis suggests that many repetitive tasks, including paper work tasks, may have a failure rate of 1.0E-3 or higher (Gertman, 1996). With between 10,000 and 45,000 subtasks being performed during an outage (Gomes, 1996), the opportunity for human error of some consequence is a realistic concern. Although a number of factors exist that can make these errors recoverable, reducing and effectively coordinating the sheer number of tasks to be performed, particularly those that are error prone, has the potential to enhance outage efficiency and safety. Additionally, outage management requires precise coordination of work groups that do not always share similar objectives. Outage

  17. Advanced Tele-operation[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M.

    1998-07-01

    Maintenance, repair, and dismantling operations in nuclear facilities have to be performed remotely when high radiation doses exclude hands-on operation, but also to minimize contamination risks and occupational doses to the operators. Computer-aided and sensor-based tele-operation enhances safety, reliability, and performance by helping the operator in difficult tasks with poor remote environmental perception. The objectives of work in this domain are to increase the scientific knowledge of the studied phenomena, to improve the interpretation of data, to improve the piloting og experimental devices during irradiation, to reveal and to understand possible unexpected phenomena occurring during irradiation. This scientific report describes the achievements for 1997 in the area of radiation tolerance for of remote-sensing, optical fibres and optical fibre sensors.

  18. Development of advanced nuclear materials - Fabrication of Zr-Nb alloy used in PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kang In; Kim, Won Baek; Lee, Chul Kyung; Choi, Kuk Sun; Kang, Dae Kyu; Seo, Chang Ryul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The following conclusions can be made from the second year research: 1. Easy control for alloying elements can be made for the following adding metals like Nb, V, Sn, Mo, Fe due to low vapor pressure. In case of Cr and Te= known to have high vapor pressure, they are controlled by adding master alloy(Zr-Cr) or quite excess of aimed composition. However, Bi was found to be very difficult to charging the certain amount into the melt. 2. Oxygen content can be adjusted by adding the Zr-10%O master alloy considering the inherent amount of oxygen in sponge zirconium. 3. The charging rod of 38 mm in diameter, 96 mm in length was made by a series of button melting, casting and vacuum welding, from this, Zr-2.5Nb ingot of 50 mm in diameter and 550 mm in length was fabricated by EB drip melting process. 4. The amount of Nb can be successfully adjusted at 2.8% with charging 15% excess. Nb as adding element is easily controlled due to high-melting -point metal and its low vapor pressure. 5. Oxygen content is not varied during remelting, casting, and drip melting, only slight change was observed in button melting stage due to uptake the desorbed gases during the melting operation. Nuclear materials in domestic nuclear power plants depend on import and this amount reaches 100 million dollars per year. The increase in demand for the development of new zirconium based alloys are expecting. All the results involving this research can be applied for the melting of reactive metals, vacuum refining and alloy design. 13 refs., 6 tabs., 10 figs., 10 ills. (author)

  19. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  20. Possibility of a gas-cooled Peltier current lead in the 200 m-class superconducting direct current transmission and distribution system of CASER-2

    Science.gov (United States)

    Kawahara, Toshio; Emoto, Masahiko; Watanabe, Hirofumi; Hamabe, Makoto; Yamaguchi, Sataro; Hikichi, Yasuo; Minowa, Masahiro

    Global energy problems should be solved quickly, and superconducting applications are highly demanded as energy saving technologies. Among them, long-distance superconducting transmission seems to be one of the most promising for energy saving by energy sharing. On the other hand, such large systems can be constructed from smaller network systems that can be enhanced by scaling up to the superconducting grid. Reducing heat leak to the low temperature end is the most important aspect of technology for practical superconducting applications, and heat leak reduction at the terminal is a key goal especially for small-length applications. At Chubu University, we have developed a 200 m-class superconducting direct current transmission and distribution system (CASER-2), in which we also used a Peltier current lead (PCL) as heat insulation at the terminal. PCL is composed of a thermoelectric material and a copper lead. In actual transmission and distribution applications, the cables are also cooled by the coolant. After the circulation, the coolant could also be used to cool the current lead. We will discuss the performance of such gas-cooled systems as the total performance of applied superconducting systems using the experimental parameters obtained in CASER-2.

  1. Buffer-gas cooling of antiprotonic helium to 1.5 to 1.7 K, and antiproton-to–electron mass ratio

    CERN Document Server

    Hori, Masaki; Sótér, Anna; Barna, Daniel; Dax, Andreas; Hayano, Ryugo; Kobayashi, Takumi; Murakami, Yohei; Todoroki, Koichi; Yamada, Hiroyuki; Horváth, Dezső; Venturelli, Luca

    2016-01-01

    Charge, parity, and time reversal (CPT) symmetry implies that a particle and its antiparticle have the same mass. The antiproton-to-electron mass ratio Embedded Image can be precisely determined from the single-photon transition frequencies of antiprotonic helium. We measured 13 such frequencies with laser spectroscopy to a fractional precision of 2.5 × 10−9 to 16 × 10−9. About 2 × 109 antiprotonic helium atoms were cooled to temperatures between 1.5 and 1.7 kelvin by using buffer-gas cooling in cryogenic low-pressure helium gas; the narrow thermal distribution led to the observation of sharp spectral lines of small thermal Doppler width. The deviation between the experimental frequencies and the results of three-body quantum electrodynamics calculations was reduced by a factor of 1.4 to 10 compared with previous single-photon experiments. From this, Embedded Image was determined as 1836.1526734(15), which agrees with a recent proton-to-electron experimental value within 8 × 10−10.

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

    Directory of Open Access Journals (Sweden)

    Christian P. Deck

    2013-01-01

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

  3. Processing of solid solution, mixed uranium/refractory metal carbides for advanced space nuclear power and propulsion systems

    Science.gov (United States)

    Knight, Travis Warren

    Nuclear thermal propulsion (NTP) and space nuclear power are two enabling technologies for the manned exploration of space and the development of research outposts in space and on other planets such as Mars. Advanced carbide nuclear fuels have been proposed for application in space nuclear power and propulsion systems. This study examined the processing technologies and optimal parameters necessary to fabricate samples of single phase, solid solution, mixed uranium/refractory metal carbides. In particular, the pseudo-ternary carbide, UC-ZrC-NbC, system was examined with uranium metal mole fractions of 5% and 10% and corresponding uranium densities of 0.8 to 1.8 gU/cc. Efforts were directed to those methods that could produce simple geometry fuel elements or wafers such as those used to fabricate a Square Lattice Honeycomb (SLHC) fuel element and reactor core. Methods of cold uniaxial pressing, sintering by induction heating, and hot pressing by self-resistance heating were investigated. Solid solution, high density (low porosity) samples greater than 95% TD were processed by cold pressing at 150 MPa and sintering above 2600 K for times longer than 90 min. Some impurity oxide phases were noted in some samples attributed to residual gases in the furnace during processing. Also, some samples noted secondary phases of carbon and UC2 due to some hyperstoichiometric powder mixtures having carbon-to-metal ratios greater than one. In all, 33 mixed carbide samples were processed and analyzed with half bearing uranium as ternary carbides of UC-ZrC-NbC. Scanning electron microscopy, x-ray diffraction, and density measurements were used to characterize samples. Samples were processed from powders of the refractory mono-carbides and UC/UC 2 or from powders of uranium hydride (UH3), graphite, and refractory metal carbides to produce hypostoichiometric mixed carbides. Samples processed from the constituent carbide powders and sintered at temperatures above the melting point of UC

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

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

    Science.gov (United States)

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

    1994-07-01

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

  6. A Condition Analysis for Advancing Exalted Level of Nuclear International Organization

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H. J.; Oh, K. B.; Chung, W. S.; Lee, K. S.; Yun, S. W.; Jeong, I.; Ko, H. S.; Lee, J. H

    2008-02-15

    This paper aimed at developing the strategy for facilitating the Korean participation as a higher staff in the nuclear related international organizations. - For that purpose, this Study focused on the following three subjects ; (1) to analyze the job situation of international organizations, (2) to analyze the background information and the procedure of appointment, (3) to suggest a facilitating strategy. - There are over 40 higher staff members including Director-General, six Deputy Director-General and 36 Directors in IAEA. Analysis shows that there are three main factors which influence the appointment of those higher position; geographical distribution, members' contributions towards the regular budget and a nation's influencing power to the international community. We can categorize into two case in aspect of the agency's appointment procedure. A director's position is generally decided through the open competition. In the case of Deputy Director-General or higher position, however, it is a practice that the Agency appoints the candidates through a unofficial procedure considering political situation. - There are twelve higher staffs in NEA. Influencing power and members' contributions towards the regular budget are the major criteria of higher staff's appointment. - It is expected that this study will contribute to the facilitation of Korean participation as a higher staff in the international organization by supplying the basic strategy.

  7. Nuclear magnetic resonance based metabolomics and liver diseases: Recent advances and future clinical applications.

    Science.gov (United States)

    Amathieu, Roland; Triba, Mohamed Nawfal; Goossens, Corentine; Bouchemal, Nadia; Nahon, Pierre; Savarin, Philippe; Le Moyec, Laurence

    2016-01-01

    Metabolomics is defined as the quantitative measurement of the dynamic multiparametric metabolic response of living systems to pathophysiological stimuli or genetic modification. It is an "omics" technique that is situated downstream of genomics, transcriptomics and proteomics. Metabolomics is recognized as a promising technique in the field of systems biology for the evaluation of global metabolic changes. During the last decade, metabolomics approaches have become widely used in the study of liver diseases for the detection of early biomarkers and altered metabolic pathways. It is a powerful technique to improve our pathophysiological knowledge of various liver diseases. It can be a useful tool to help clinicians in the diagnostic process especially to distinguish malignant and non-malignant liver disease as well as to determine the etiology or severity of the liver disease. It can also assess therapeutic response or predict drug induced liver injury. Nevertheless, the usefulness of metabolomics is often not understood by clinicians, especially the concept of metabolomics profiling or fingerprinting. In the present work, after a concise description of the different techniques and processes used in metabolomics, we will review the main research on this subject by focusing specifically on in vitro proton nuclear magnetic resonance spectroscopy based metabolomics approaches in human studies. We will first consider the clinical point of view enlighten physicians on this new approach and emphasis its future use in clinical "routine".

  8. Advanced characterisation of encapsulated lipid powders regarding microstructure by time domain-nuclear magnetic resonance.

    Science.gov (United States)

    Linke, Annika; Anzmann, Theresa; Weiss, Jochen; Kohlus, Reinhard

    2017-03-15

    Encapsulation is an established technique to protect sensitive materials from environmental stress. In order to understand the physical protection mechanism against oxidation, knowledge about the powder microstructure is required. Time domain-nuclear magnetic resonance (TD-NMR) has the potential to determine the surface oil (SO) and droplet size distribution by relaxation and restricted self-diffusion, respectively. The amount of SO, the retention and encapsulation efficiency are determined based on a lipid balance. The oil load of the initial powder and after SO removal is measured by TD-NMR. The results correlate with gravimetric and photometric references. The oil droplet size obtained by TD-NMR correlates well with static light scattering. The diameter of droplets in emulsions and dried powder both measured by TD-NMR, correlates (r = 0.998), implying that oil droplets embedded in a solid matrix can be measured. Summarising, TD-NMR allows analysis of the microstructure of encapsulated lipid powders, in a rapid, simple and non-destructive way.

  9. First-principles modelling of radiation defects in advanced nuclear fuels

    Science.gov (United States)

    Kotomin, E. A.; Gryaznov, D.; Grimes, R. W.; Parfitt, D.; Zhukovskii, Yu. F.; Mastrikov, Yu. A.; Van Uffelen, P.; Rondinella, V. V.; Konings, R. J. M.

    2008-06-01

    We present and discuss the results of the first-principles calculations of Frenkel defects and O impurities in uranium mononitride (UN) perspective for fast reactor nuclear fuels. Special attention is paid to the calculation of defect migration energies. We demonstrate that the interstitialcy mechanism (with the formation of a N-N dumbbell along the [1 1 1] axis) is energetically more favorable than the direct [1 0 0] hops. As a result, for the interstitial N ions we predict a diffusion mechanism similar to that known in isostructural fcc materials with a different chemical nature (KCl, MgO). The calculated effective N charge considerably depends on the ion position and environment (a host lattice site, interstitial or saddle point) which strongly limits the applicability of classical defect modelling based on formal invariant charges. Lastly, the calculated migration energy for the interstitial impurity O ions is quite low (2.84 eV), which indicates their high mobility and ability for reactions with other defects.

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

    Science.gov (United States)

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

    2013-01-01

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

  11. Advancing Inverse Sensitivity/Uncertainty Methods for Nuclear Fuel Cycle Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arbanas, Goran [ORNL; Williams, Mark L [ORNL; Leal, Luiz C [ORNL; Dunn, Michael E [ORNL; Khuwaileh, Bassam A. [North Carolina State University; Wang, C [North Carolina State University; Abdel-Khalik, Hany [North Carolina State University

    2015-01-01

    The inverse sensitivity/uncertainty quantification (IS/UQ) method has recently been implemented in the Inverse Sensitivity/UnceRtainty Estimiator (INSURE) module of the AMPX system [1]. The IS/UQ method aims to quantify and prioritize the cross section measurements along with uncertainties needed to yield a given nuclear application(s) target response uncertainty, and doing this at a minimum cost. Since in some cases the extant uncertainties of the differential cross section data are already near the limits of the present-day state-of-the-art measurements, requiring significantly smaller uncertainties may be unrealistic. Therefore we have incorporated integral benchmark experiments (IBEs) data into the IS/UQ method using the generalized linear least-squares method, and have implemented it in the INSURE module. We show how the IS/UQ method could be applied to systematic and statistical uncertainties in a self-consistent way. We show how the IS/UQ method could be used to optimize uncertainties of IBEs and differential cross section data simultaneously.

  12. DEVELOPMENT OF CERAMIC WASTE FORMS FOR AN ADVANCED NUCLEAR FUEL CYCLE

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.; Billings, A.; Brinkman, K.; Fox, K.

    2010-11-30

    A series of ceramic waste forms were developed and characterized for the immobilization of a Cesium/Lanthanide (CS/LN) waste stream anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3} and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores and other minor metal titanate phases. Three fabrication methodologies were used, including melting and crystallizing, pressing and sintering, and Spark Plasma Sintering (SPS), with the intent of studying phase evolution under various sintering conditions. X-Ray Diffraction (XRD) and Scanning Electron Microscopy coupled with Energy Dispersive Spectroscopy (SEM/EDS) results showed that the partitioning of the waste elements in the sintered materials was very similar, despite varying stoichiometry of the phases formed. Identification of excess Al{sub 2}O{sub 3} via XRD and SEM/EDS in the first series of compositions led to a Phase II study, with significantly reduced Al{sub 2}O{sub 3} concentrations and increased waste loadings. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD. Chemical composition measurements showed no significant issues with meeting the target compositions. However, volatilization of Cs and Mo was identified, particularly during melting, since sintering of the pressed pellets and SPS were performed at lower temperatures. Partitioning of some of the waste components was difficult to determine via XRD. SEM/EDS mapping showed that those elements, which were generally present in small concentrations, were well distributed throughout the waste forms.

  13. Advanced biological treatment of aqueous effluent from the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Pitt, Jr, W W; Hancher, C W; Patton, B D; Shumate, II, S E

    1980-01-01

    Many of the processing steps in the nuclear fuel cycle generate aqueous effluent streams bearing contaminants that can, because of their chemical or radiological properties, pose an environmental hazard. Concentration of such contaminants must be reduced to acceptable levels before the streams can be discharged to the environment. Two classes of contaminants, nitrates and heavy metals, are addressed in this study. Specific techniques aimed at the removal of nitrates and radioactive heavy metals by biological processes are being developed, tested, and demonstrated. Although cost comparisons between biological processes and current treatment methods will be presented, these comparisons may be misleading because biological processes yield environmentally better end results which are difficult to price. The fluidized-bed biological denitrification process is an environmentally acceptable and economically sound method for the disposal of nonreusable sources of nitrate effluents. A very high denitrification rate can be obtained in a FBR as the result of a high concentration of denitrification bacteria in the bioreactor and the stagewise operation resulting from plug flow in the reactor. The overall denitrification rate in an FBR ranges from 20- to 100-fold greater than that observed for an STR bioreactor. It has been shown that the system can be operated using Ca/sup 2 +/, Na/sup +/, or NH/sub 4//sup +/ cations at nitrate concentrations up to 1 g/liter without inhibition. Biological sorption of uranium and other radionuclides (particularly the actinides) from dilute aqueous waste streams shows considerable promise as a means of recovering these valuable resources and reducing the environmental impact, however, further development efforts are required.

  14. Americium/Lanthanide Separations in Alkaline Solutions for Advanced Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Goff, George S. [Los Alamos National Laboratory; Long, Kristy Marie [Los Alamos National Laboratory; Reilly, Sean D. [Los Alamos National Laboratory; Jarvinen, Gordon D. [Los Alamos National Laboratory; Runde, Wolfgang H. [Los Alamos National Laboratory

    2012-06-11

    Project goals: Can used nuclear fuel be partitioned by dissolution in alkaline aqueous solution to give a solution of uranium, neptunium, plutonium, americium and curium and a filterable solid containing nearly all of the lanthanide fission products and certain other fission products? What is the chemistry of Am/Cm/Ln in oxidative carbonate solutions? Can higher oxidation states of Am be stabilized and exploited? Conclusions: Am(VI) is kinetically stable in 0.5-2.0 M carbonate solutions for hours. Aliquat 336 in toluene has been successfully shown to extract U(VI) and Pu(VI) from carbonate solutions. (Stepanov et al 2011). Higher carbonate concentration gives lower D, SF{sub U/Eu} for = 4 in 1 M K{sub 2}CO{sub 3}. Experiments with Am(VI) were unsuccessful due to reduction by the organics. Multiple sources of reducing organics...more optimization. Reduction experiments of Am(VI) in dodecane/octanol/Aliquat 336 show that after 5 minutes of contact, only 30-40% of the Am(VI) has been reduced. Long enough to perform an extraction. Shorter contact times, lower T, and lower Aliquat 336 concentration still did not result in any significant extraction of Am. Anion exchange experiments using a strong base anion exchanger show uptake of U(VI) with minimal uptake of Nd(III). Experiments with Am(VI) indicate Am sorption with a Kd of 9 (10 minute contact) but sorption mechanism is not yet understood. SF{sub U/Nd} for = 7 and SF{sub U/Eu} for = 19 after 24 hours in 1 M K{sub 2}CO{sub 3}.

  15. Development of advanced risk informed asset management tool based on system dynamics approach for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Gyoung Cheol

    2007-02-15

    In the competitive circumstance of electricity industry, the economic efficiency of electricity generation facility is the most important factor to increase their competitiveness. For nuclear power plant (NPP), safety is also an essential factor. Over fast several years, efforts for development of safety concerned and financial asset maximizing method, process and tools have been continued internationally and Risk-Informed Asset Management (RIAM) methodology is suggested by Electric Power Research Institute (EPRI). This RIAM methodology is expected to provide plant operators with a project prioritization and life cycle management planning tool for making long-term maintenance plans, guiding plant budgeting, and determining the sensitivity of a plant's economic risk to the reliability and availability of system, structure, and components (SSC), as well as other technical and economic parameters. The focus of this study is to develop model that help us to resource allocation, to find what effect such allocations on the plant economic and safety performance. Detailed research process for this goal is as follow; First step for development of advanced RIAM model is to review for current RIAM model of EPRI. This part describes the overall RIAM methodology including its conceptual model, implementation process, modular approach etc. Second step is to perform feasibility study for current EPRI's RIAM model with case study. This part shows the result of feasibility study for current RIAM method by case study and discussion for result. Finally, concept of Advanced RIAM model is developed based on system dynamics approach and parameter relationship is formulated. In advanced RIAM Model, Identification of scheduled maintenance effect on other parameters and the relationship between PM Activity and failure rate is most important factor. In this study, these relationships are formulated based on system dynamics approach. Creations of these modeling tool using Vensim

  16. 75 FR 75641 - Advance Notification to Native American Tribes of Transportation of Certain Types of Nuclear Waste

    Science.gov (United States)

    2010-12-06

    ... Tribes of Transportation of Certain Types of Nuclear Waste AGENCY: Nuclear Regulatory Commission. ACTION...'s designee, of certain shipments of irradiated reactor fuel and certain nuclear waste passing... notification to Native American Tribes of transportation of certain types of nuclear waste (64 FR...

  17. A review of modern advances in analyses and applications of single-phase natural circulation loop in nuclear thermal hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Basu, Dipankar N., E-mail: dipankar.n.basu@gmail.com [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati 781039 (India); Bhattacharyya, Souvik; Das, P.K. [Department of Mechanical Engineering, Indian Institute of Technology Kharagpur, Kharagpur 721302 (India)

    2014-12-15

    Highlights: • Comprehensive review of state-of-the-art on single-phase natural circulation loops. • Detailed discussion on growth in solar thermal system and nuclear thermal hydraulics. • Systematic development in scaling methodologies for fabrication of test facilities. • Importance of numerical modeling schemes for stability assessment using 1-D codes. • Appraisal of current trend of research and possible future directions. - Abstract: A comprehensive review of single-phase natural circulation loop (NCL) is presented here. Relevant literature reported since the later part of 1980s has been meticulously surveyed, with occasional obligatory reference to a few pioneering studies originating prior to that period, summarizing the key observations and the present trend of research. Development in the concept of buoyancy-induced flow is discussed, with introduction to flow initiation in an NCL due to instability. Detailed discussion on modern advancement in important application areas like solar thermal systems and nuclear thermal hydraulics are presented, with separate analysis for various reactor designs working on natural circulation. Identification of scaling criteria for designing lab-scale experimental facilities has gone through a series of modification. A systematic analysis of the same is presented, considering the state-of-the-art knowledge base. Different approaches have been followed for modeling single-phase NCLs, including simplified Lorenz system mostly for toroidal loops, 1-D computational modeling for both steady-state and stability characterization and 3-D commercial system codes to have a better flow visualization. Methodical review of the relevant studies is presented following a systematic approach, to assess the gradual progression in understanding of the practical system. Brief appraisal of current research interest is reported, including the use of nanofluids for fluid property augmentation, marine reactors subjected to rolling waves

  18. 球床高温气冷堆闭式循环特性%Characteristics of closed fuel cycles in the pebble bed high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    位金锋; 孙玉良; 李富

    2012-01-01

    The reuse of uranium and plutonium from high temperature gas-cooled reactor(HTGR) spent fuel will improve resource usage and minimize waste.The characteristics of different closed fuel cycles were studied here for uranium and plutonium recycled from 250 MWth high-temperature gas-cooled reactor pebble-bed-module(HTR-PM) spent fuel from a U-Pu fueled core.PuO2 and MOX fuel elements using recycled plutonium and uranium were then used in new PuO2 or MOX fueled cores with the same geometry as the original reactor.PuO2 from LWR spent fuel was also evaluated.The characteristics of the fuel utilization and transuranic incineration in these closed fuel cycles were studied with the VSOP program.The natural uranium utilization closed fuel for these closed fuel cycle is increased by 6%,8% and 20%,while the plutonium burn rates are 40%,41% and 63%,respectively.Thus,these HTGR closed fuel cycles can effectively burn plutonium isotopes and increase natural uranium utilization.%从提高天然铀利用率和改进废物管理方面考虑,研究球床高温气冷堆乏燃料中铀钚的再利用和不同闭式燃料循环的特性。在250MW热功率球床模块式高温气冷堆示范电站铀钚循环的乏燃料中提取铀和钚为核燃料,设计了PuO2和混合氧化物(MOX)燃料元件,将新设计的燃料元件重新装入与示范电站有同样结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。还研究了基于轻水堆级钚的燃料循环。采用了高温气冷堆物理设计程序VSOP,研究了高温气冷堆不同闭式循环的燃料利用和超铀元素焚烧特性。不同闭式循环钚消耗率分别为50%、46%和71%,天然铀的电利用率分别提高了6%、8%和20%。结果表明:高温气冷堆闭式燃料循环能有效焚烧钚同位素,适度提高天然铀的利用率。

  19. Deterministic and risk-informed approaches for safety analysis of advanced reactors: Part II, Risk-informed approaches

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Inn Seock, E-mail: innseockkim@gmail.co [ISSA Technology, 21318 Seneca Crossing Drive, Germantown, MD 20876 (United States); Ahn, Sang Kyu; Oh, Kyu Myung [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)

    2010-05-15

    Technical insights and findings from a critical review of deterministic approaches typically applied to ensure design safety of nuclear power plants were presented in the companion paper of Part I included in this issue. In this paper we discuss the risk-informed approaches that have been proposed to make a safety case for advanced reactors including Generation-IV reactors such as Modular High-Temperature Gas-cooled Reactor (MHTGR), Pebble Bed Modular Reactor (PBMR), or Sodium-cooled Fast Reactor (SFR). Also considered herein are a risk-informed safety analysis approach suggested by Westinghouse as a means to improve the conventional accident analysis, together with the Technology Neutral Framework recently developed by the US Nuclear Regulatory Commission as a high-level regulatory infrastructure for safety evaluation of any type of reactor design. The insights from a comparative review of various deterministic and risk-informed approaches could be usefully used in developing a new licensing architecture for enhanced safety of evolutionary or advanced plants.

  20. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Science.gov (United States)

    Ariani, Menik; Satya, Octavianus Cakra; Monado, Fiber; Su'ud, Zaki; Sekimoto, Hiroshi

    2016-03-01

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on "Region-8" and "Region-10" core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  1. The prospect of uranium nitride (UN-PuN) fuel for 25- 100MWe gas cooled fast reactor long life without refuelling

    Science.gov (United States)

    Syarifah, R. D.; Su'ud, Z.; Basar, K.; Irwanto, D.

    2016-11-01

    The prospect of uranium nitride (UN-PuN) fuel for 25-100MWe Gas Cooled Fast Reactor has been done. This research use helium coolant which has low neutron moderation, chemical inert and single phase. This study use natural uranium and plutonium. Plutonium taken from spent fuel of LWR (Light Water Reactor). So, it can reduced spent fuel in the world. The calculation use SRAC2006 and JENDL 4.0 for the data libraries. First, we calculate PIJ for fuel pin cell calculation and CITATION for core calculation. The reflector radial-axial width is 50 cm. The variation of fuel fraction is 40% until 65%, cladding 10%, and moderator 25% up to 50%. The variation of the power is 75-300 MWth (25-100 MWe). The calculation of survey parameter has been done. The variation of percentage plutonium is 7% up to 13%. We have optimum k-eff value in percentage of plutonium 11%. The high powers cause k-eff value high too. Second, the core configuration divided by three variation fuel (F1, F2, and F3). F1 is located in the central core, F2 middle core and F3 outer core. The variation percentage Plutonium for fuel F1:F2:F3 = 8%:10%:12%. The increasing power level make the burn up level increase. All case can reach burn up time plus than 20 years. The thermal powers increase cause the peak power density increase. The power 150 MWth, 225 MWth, and 300 MWth have excess reactivity (%Ak/k) less than 2%.

  2. Boron Depletion in High-Temperature Gas-Cooled Reactor%杂质硼在高温气冷堆中的燃耗特性

    Institute of Scientific and Technical Information of China (English)

    赵晶; 李富; 魏春琳

    2012-01-01

    There is a small quantity of boron as impurity in the core and graphite reflector of pebble bed high-temperature gas-cooled reactor (HTR). Boron and its change along depletion have influence on the reactivity of the reactor. The depletion characteristics of boron were calculated for each batch of fuel element along its operation history for the multi-pass pebble bed core, and for each region of graphite reflector. The reactivity worth of boron and its change along depletion were calculated with the perturbation theory. According to the analysis, the boron is depleted rapidly, therefore the influence on the reactivity also reduces rapidly.%在球床式高温气冷堆的堆芯和石墨反射层中,不可避免地含有少量杂质硼.硼杂质的存在及其燃耗会对反应堆的反应性产生影响.对于多次通过的球床堆芯,根据燃料元件的运行历史计算所有元件的硼燃耗,对于中子注量率差别较大的反射层,分区计算了硼燃耗.再采用微扰理论,计算燃耗过程中硼反应性价值的变化.计算结果表明,硼杂质燃耗很快,因此,硼杂质对反应性的影响降低很快.

  3. Time-Dependent Neutronic Analysis of a Power-Flattened Gas Cooled Accelerator Driven System Fuelled with Thorium, Uranium, Plutonium, and Curium Dioxides TRISO Particles

    Directory of Open Access Journals (Sweden)

    Gizem Bakır

    2016-01-01

    Full Text Available This study presents the power flattening and time-dependent neutronic analysis of a conceptual helium gas cooled Accelerator Driven System (ADS loaded with TRISO (tristructural-isotropic fuel particles. Target material is lead-bismuth eutectic (LBE. ThO2, UO2, PuO2, and CmO2 TRISO particles are used as fuel. PuO2 and CmO2 fuels are extracted from PWR-MOX spent fuel. Subcritical core is radially divided into 10 equidistant subzones in order to flatten the power produced in the core. Tens of thousands of these TRISO fuel particles are embedded in the carbon matrix fuel pebbles as five different cases. The high-energy Monte Carlo code MCNPX 2.7 with the LA150 library is used for the neutronic calculations. Time-dependent burnup calculations are carried out for thermal fission power (Pth of 1000 MW using the BURN card. The energy gain of the ADS is in the range of 99.98–148.64 at the beginning of a cycle. Furthermore, the peak-to-average fission power density ratio is obtained between 1.021 and 1.029 at the beginning of the cycle. These ratios show a good quasi-uniform power density for each case. Furthermore, up to 155.1 g 233U and 103.6 g 239Pu per day can be produced. The considered system has a high neutronic capability in terms of energy multiplication, fissile breeding, and spent fuel transmutation with thorium utilization.

  4. SBBN 2010: 7. Congress of the Brazilian Society of Nuclear Biosciences. Radiations in biosciences: advances and trends; SBBN 2010: 7. Congresso da Sociedade Brasileira de Biociencias Nucleares. Radiacoes em biociencias: avancos e perspectivas

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    Advance and new perspectives related to the use of ionizing and no ionizing radiations in nuclear biosciences are presented. Multidisciplinary approach, including radiopharmacy, radioprotection and dosimetry, cytogenetic, biosafety, radioecology, environmental toxicology are studied. Topics of Nuclear medicine, radiotherapy and image diagnosis, such as computerized tomography, PET scan, phantoms, biomedical radiography, are reported. Use of radioisotopes, evaluation of radiation dose rates, radiation dose distribution, radiation monitoring is considered. Environmental impact of radiation are also in human beings, animals and for several purposes are analyzed. (MAC)

  5. Advanced characterization of pores and fractures in coals by nuclear magnetic resonance and X-ray computed tomography

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    This paper demonstrates capabilities of low-field nuclear magnetic resonance (NMR) and microfocus X-ray computed tomography (μCT) in advanced, nondestructive, and quantitative characterization of pore types, producible porosity, pore structure, and spatial disposition of pore-fractures in coals. Results show that the NMR transverse relaxation time (T2) at 0.5–2.5, 20–50, and >100 ms correspond to pores of <0.1 μm, >0.1 μm, and fractures, respectively. A much higher T2 spectrum peak reflects a much better development of pores (or fractures) corresponding to the T2, and vice versa. Three basic components in coals, i.e., the pores (or fractures), coal matrix, and minerals have their distinctive range of CT numbers. Among these, the CT number of pores is commonly less than 600 HU. The producible porosity, which is a determination of permeability, can be calculated by T2 cutoff value (T2C) of coal NMR. The coal pore structure can be efficiently estimated by the newly proposed "T2C based model". Finally, μCT scan was proven capable of modeling and spatial visualization of pores and fractures.

  6. AIDA: A 16-channel amplifier ASIC to read out the advanced implantation detector array for experiments in nuclear decay spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Braga, D. [STFC Rutherford Appleton Laboratory, Didcot, OX11 0QX (United Kingdom); Coleman-Smith, P. J. [STFC Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Davinson, T. [Dept. of Physics and Astronomy, Univ. of Edinburgh, Edinburgh EH9 3JZ (United Kingdom); Lazarus, I. H. [STFC Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Page, R. D. [Dept. of Physics, Univ. of Liverpool, Oliver Lodge Laboratory, Liverpool L69 7ZE (United Kingdom); Thomas, S. [STFC Rutherford Appleton Laboratory, Didcot, OX11 0QX (United Kingdom)

    2011-07-01

    We have designed a read-out ASIC for nuclear decay spectroscopy as part of the AIDA project - the Advanced Implantation Detector Array. AIDA will be installed in experiments at the Facility for Antiproton and Ion Research in GSI, Darmstadt. The AIDA ASIC will measure the signals when unstable nuclei are implanted into the detector, followed by the much smaller signals when the nuclei subsequently decay. Implant energies can be as high as 20 GeV; decay products need to be measured down to 25 keV within just a few microseconds of the initial implants. The ASIC uses two amplifiers per detector channel, one covering the 20 GeV dynamic range, the other selectable over a 20 MeV or 1 GeV range. The amplifiers are linked together by bypass transistors which are normally switched off. The arrival of a large signal causes saturation of the low-energy amplifier and a fluctuation of the input voltage, which activates the link to the high-energy amplifier. The bypass transistors switch on and the input charge is integrated by the high-energy amplifier. The signal is shaped and stored by a peak-hold, then read out on a multiplexed output. Control logic resets the amplifiers and bypass circuit, allowing the low-energy amplifier to measure the subsequent decay signal. We present simulations and test results, demonstrating the AIDA ASIC operation over a wide range of input signals. (authors)

  7. Recent advances in biosynthetic modeling of nitric oxide reductases and insights gained from nuclear resonance vibrational and other spectroscopic studies

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Saumen; Reed, Julian; Sage, Timothy; Branagan, Nicole C.; Petrik, Igor D.; Miner, Kyle D.; Hu, Michael Y.; Zhao, Jiyong; Alp, E. Ercan; Lu, Yi

    2015-10-05

    This Forum Article focuses on recent advances in structural and spectroscopic studies of biosynthetic models of nitric oxide reductases (NORs). NORs are complex metalloenzymes found in the denitrification pathway of Earth's nitrogen cycle where they catalyze the proton-dependent twoelectron reduction of nitric oxide (NO) to nitrous oxide (N2O). While much progress has been made in biochemical and biophysical studies of native NORs and their variants, a. clear mechanistic understanding of this important metalloenzyme related to its function is still elusive. We report herein UV vis and nuclear resonance vibrational spectroscopy (NRVS) studies of mononitrosylated intermediates of the NOR reaction of a biosynthetic model. The ability to selectively substitute metals at either heme or nonheme metal sites allows the introduction of independent 57Fe probe atoms at either site, as well as allowing the preparation of analogues of stable reaction intermediates by replacing either metal with a redox inactive metal. Together with previous structural and spectroscopic results, we summarize insights gained from studying these biosynthetic models toward understanding structural features responsible for the NOR activity and its mechanism. As a result, the outlook on NOR modeling is also discussed, with an emphasis on the design of models capable of catalytic turnovers designed based on close mimics of the secondary coordination sphere of native NORs.

  8. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

    Energy Technology Data Exchange (ETDEWEB)

    Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

    2011-02-01

    This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

  9. 新型气冷式射频消融系统的控制模式%Control mode of a new gas-cooled controlled radio frequency ablation system

    Institute of Scientific and Technical Information of China (English)

    王玉凯; 谷雪莲; 常兆华; 赵庆孝

    2011-01-01

    BACKGROUND: Radio frequency ablation (RFA) has widely utilized in the treatment for tumor with minimally invasion. However,the current RFA has disadvantages of little volume of coagulation.OBJECTIVE: To design a new gas-cooled RFA controlled mode to increase the extent of coagulation.METHODS: A new gas-cooled RFA controlled mode was researched based on existing gas-cooled RFA system and characteristics of gas-cooled electrode. That is, the new mode was by freeing power of gas-cooled to control the range of RFA under temperature-controlled. The control system adopted master machine and slave machine architecture, and the temperature controlling used improved PID control algorithm.RESULTS AND CONCLUSION: The temperature control results showed that improved PID control algorithm could control temperature precisely, with a fast reaction velocity. Gas-cooled radio frequency power and cooling power showed that, freeing power of gas-cooled can better regulate the output power of RFA to achieve aim of controlling RFA range.%背景:射频消融作为一种有前途的微创肿瘤治疗方法应用越来越广泛,但是现今的射频消融系统还普遍存在消融范围小的问题.目的:设计一种新型的气冷式射频消融电极的控制模式,增大射频消融的范围.方法:在气冷式射频仪的基础上,根据气冷式射频电极的特点,设计了一种新型的气冷式射频消融的控制模式,即在恒温的方式下通过气冷式电极的冷冻功率来控制射频消融的范围.控制系统采用上位机、下位机的结构,温度控制算法采用改进的PID控制算法.结果与结论:通过温度控制实验证明,改进的PID控制算法能精确地控制温度,反应速度快,能很好地应用于该控制模式.气冷式射频功率与制冷功率关系实验结果表明,通过气冷式冷却电极的功率能很好地调节射频消融的射频输出功率,达到控制消融范围的目的.

  10. Evaluation of Alternate Materials for Coated Particle Fuels for the Gas-Cooled Fast Reactor. Laboratory Directed Research and Development Program FY 2006 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; Karen Wright; Jian Gan; David Petti; Todd Allen; Jake Blanchard

    2006-09-01

    Candidate ceramic materials were studied to determine their suitability as Gas-Cooled Fast Reactor particle fuel coatings. The ceramics examined in this work were: TiC, TiN, ZrC, ZrN, AlN, and SiC. The studies focused on (i) chemical reactivity of the ceramics with fission products palladium and rhodium, (ii) the thermomechanical stresses that develop in the fuel coatings from a variety of causes during burnup, and (iii) the radiation resiliency of the materials. The chemical reactivity of TiC, TiN, ZrC, and ZrN with Pd and Rh were all found to be much lower than that of SiC. A number of important chemical behaviors were observed at the ceramic-metal interfaces, including the formation of specific intermetallic phases and a variation in reaction rates for the different ceramics investigated. Based on the data collected in this work, the nitride ceramics (TiN and ZrN) exhibit chemical behavior that is characterized by lower reaction rates with Pd and Rh than the carbides TiC and ZrC. The thermomechanical stresses in spherical fuel particle ceramic coatings were modeled using finite element analysis, and included contributions from differential thermal expansion, fission gas pressure, fuel kernel swelling, and thermal creep. In general the tangential stresses in the coatings during full reactor operation are tensile, with ZrC showing the lowest values among TiC, ZrC, and SiC (TiN and ZrN were excluded from the comprehensive calculations due to a lack of available materials data). The work has highlighted the fact that thermal creep plays a critical role in the development of the stress state of the coatings by relaxing many of the stresses at high temperatures. To perform ion irradiations of sample materials, an irradiation beamline and high-temperature sample irradiation stage was constructed at the University of Wisconsin’s 1.7MV Tandem Accelerator Facility. This facility is now capable of irradiating of materials to high dose while controlling sample temperature

  11. Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).

    Energy Technology Data Exchange (ETDEWEB)

    Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

    2010-09-01

    This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

  12. Application of Spatial Data Modeling Systems, Geographical Information Systems (GIS), and Transportation Routing Optimization Methods for Evaluating Integrated Deployment of Interim Spent Fuel Storage Installations and Advanced Nuclear Plants

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Gary T [ORNL; Belles, Randy [ORNL; Cetiner, Sacit M [ORNL; Howard, Rob L [ORNL; Liu, Cheng [ORNL; Mueller, Don [ORNL; Omitaomu, Olufemi A [ORNL; Peterson, Steven K [ORNL; Scaglione, John M [ORNL

    2012-06-01

    The objective of this siting study work is to support DOE in evaluating integrated advanced nuclear plant and ISFSI deployment options in the future. This study looks at several nuclear power plant growth scenarios that consider the locations of existing and planned commercial nuclear power plants integrated with the establishment of consolidated interim spent fuel storage installations (ISFSIs). This research project is aimed at providing methodologies, information, and insights that inform the process for determining and optimizing candidate areas for new advanced nuclear power generation plants and consolidated ISFSIs to meet projected US electric power demands for the future.

  13. Role of Nuclear Grade Graphite in Oxidation in Modular HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Willaim Windes; G. Strydom; J. Kane; R. Smith

    2014-11-01

    The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of core environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.

  14. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    Science.gov (United States)

    2010-03-05

    with uranium to make mixed-oxide ( MOX ) fuel, in which the 239Pu largely substitutes for 235U. Two French reprocessing plants at La Hague can each...and France also have older plants to reprocess gas-cooled reactor fuel, and India has a 275-ton plant.53 About 200 metric tons of MOX fuel is used...to make MOX fuel for today’s nuclear power plants are modest. Existing commercial light water reactors use ordinary water to slow down, or “moderate

  15. The potential predictive role of nuclear NHERF1 expression in advanced gastric cancer patients treated with epirubicin/oxaliplatin/capecitabine first line chemotherapy

    Science.gov (United States)

    Mangia, Anita; Caldarola, Lucia; Dell'Endice, Stefania; Scarpi, Emanuela; Saragoni, Luca; Monti, Manlio; Santini, Daniele; Brunetti, Oronzo; Simone, Giovanni; Silvestris, Nicola

    2015-01-01

    Cellular resistance in advanced gastric cancer (GC) might be related to function of multidrug resistance (MDR) proteins. The adaptor protein NHERF1 (Na+/H+ exchanger regulatory factor) is an important player in cancer progression for a number of solid malignancies, even if its role to develop drug resistance remains uncertain. Herein, we aimed to analyze the potential association between NHERF1 expression and P-gp, sorcin and HIF-1α MDR-related proteins in advanced GC patients treated with epirubicin/oxaliplatin/capecitabine (EOX) chemotherapy regimen, and its relation to response. Total number of 28 untreated patients were included into the study. Expression and subcellular localization of all proteins were assessed by immunohistochemistry on formalin-fixed paraffin embedded tumor samples. We did not found significant association between NHERF1 expression and the MDR-related proteins. A trend was observed between positive cytoplasmic NHERF1 (cNHERF1) expression and negative nuclear HIF-1α (nHIF-1α) expression (68.8% versus 31.3% respectively, P = 0.054). However, cytoplasmic P-gp (cP-gp) expression was positively correlated with both cHIF-1α and sorcin expression (P = 0.011; P = 0.002, respectively). Interestingly, nuclear NHERF1 (nNHERF1) staining was statistically associated with clinical response. In detail, 66.7% of patients with high nNHERF1 expression had a disease control rate, while 84.6% of subjects with negative nuclear expression of the protein showed progressive disease (P = 0.009). Multivariate analysis confirmed a significant correlation between nNHERF1 and clinical response (OR 0.06, P = 0.019). These results suggest that nuclear NHERF1 could be related to resistance to the EOX regimen in advanced GC patients, identifying this marker as a possible independent predictive factor. PMID:26126066

  16. Advanced fuel cycle cost estimation model and its cost estimation results for three nuclear fuel cycles using a dynamic model in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sungki, E-mail: sgkim1@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ko, Wonil [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Youn, Saerom; Gao, Ruxing [University of Science and Technology, 217 Gajungro, Yuseong-gu, Daejeon 305-350 (Korea, Republic of); Bang, Sungsig, E-mail: ssbang@kaist.ac.kr [Korea Advanced Institute of Science and Technology, Department of Business and Technology Management, 291 Deahak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2015-11-15

    Highlights: • The nuclear fuel cycle cost using a new cost estimation model was analyzed. • The material flows of three nuclear fuel cycle options were calculated. • The generation cost of once-through was estimated to be 66.88 mills/kW h. • The generation cost of pyro-SFR recycling was estimated to be 78.06 mills/kW h. • The reactor cost was identified as the main cost driver of pyro-SFR recycling. - Abstract: The present study analyzes advanced nuclear fuel cycle cost estimation models such as the different discount rate model and its cost estimation results. To do so, an analysis of the nuclear fuel cycle cost of three options (direct disposal (once through), PWR–MOX (Mixed OXide fuel), and Pyro-SFR (Sodium-cooled Fast Reactor)) from the viewpoint of economic sense, focusing on the cost estimation model, was conducted using a dynamic model. From an analysis of the fuel cycle cost estimation results, it was found that some cost gap exists between the traditional same discount rate model and the advanced different discount rate model. However, this gap does not change the priority of the nuclear fuel cycle option from the viewpoint of economics. In addition, the fuel cycle costs of OT (Once-Through) and Pyro-SFR recycling based on the most likely value using a probabilistic cost estimation except for reactor costs were calculated to be 8.75 mills/kW h and 8.30 mills/kW h, respectively. Namely, the Pyro-SFR recycling option was more economical than the direct disposal option. However, if the reactor cost is considered, the economic sense in the generation cost between the two options (direct disposal vs. Pyro-SFR recycling) can be changed because of the high reactor cost of an SFR.

  17. Novel fabrication of silicon carbide based ceramics for nuclear applications

    Science.gov (United States)

    Singh, Abhishek Kumar

    Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These materials include refractory alloys based on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as SiC--SiCf; carbon--carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor components is necessary for improved efficiency. Improving thermal conductivity of the fuel can lower the center-line temperature and, thereby, enhance power production capabilities and reduce the risk of premature fuel pellet failure. Crystalline silicon carbide has superior characteristics as a structural material from the viewpoint of its thermal and mechanical properties, thermal shock resistance, chemical stability, and low radioactivation. Therefore, there have been many efforts to develop SiC based composites in various forms for use in advanced energy systems. In recent years, with the development of high yield preceramic precursors, the polymer infiltration and pyrolysis (PIP) method has aroused interest for the fabrication of ceramic based materials, for various applications ranging from disc brakes to nuclear reactor fuels. The pyrolysis of preceramic polymers allow new types of ceramic materials to be processed at relatively low temperatures. The raw materials are element-organic polymers whose composition and architecture can be tailored and varied. The primary focus of this study is to use a pyrolysis based process to fabricate a host of novel silicon carbide-metal carbide or oxide composites, and to synthesize new materials based on mixed-metal silicocarbides that cannot be processed using conventional techniques. Allylhydridopolycarbosilane (AHPCS), which is an organometal polymer, was used as the precursor for silicon carbide. Inert gas pyrolysis of AHPCS produces near-stoichiometric amorphous

  18. How operational Advanced-DInSAR Analysis can improve knowledge on natural and anthropogenic deformations for Nuclear Power Plant areas

    Science.gov (United States)

    Vollrath, Andreas; Zucca, Francesco; Stramondo, Salvatore; Bignami, Christian; Roeder, Johannes

    2015-04-01

    The application of Advanced Differential Interferometric Synthetic Aperture Radar (A-DINSAR) techniques has strongly emerged in the last two decades and became an important part in georelated fields. State-of-the-art A-DInSAR methods, such as Persistent Scatterer Interferometry (PSI) or the Small BASeline (SBAS) approach have demonstrated their usefulness in monitoring urban areas and single buildings, up to critical infrastructures. Combined with additional data from GPS networks or levelling, it could prove its large potential for an operational, cost-effective mapping of surface deformations. Given a reasonable amount of images, changes in surface deformation can be detected down to 1 mm/y. Compared to point-wise field measurements it offers a spatially consistent mapping approach from local to regional scales. In this review we want to provide a synopsis how A-DInSAR can be utilized in the framework of Nuclear Power Plant safety. Indeed, A-DInSAR is able to provide a detailed spatial analysis of slow movements occurring at NPP structures directly, as well as within the surrounding areas of the NPPs. Different phenomena of surface motion can be subject of such a monitoring. Natural causes, like active tectonics and terrain instability of slope which lead to landslides, as well as human-induced subsidence phenomena due to heavy construction or water pumping can be detected. We start by presenting techniques to determine the feasibility of the analysis for a given area and show its limitations. Then we propose a short insight into state-of-the-art studies where landslides, interseismic and human-induced deformation of the surface were mapped by A-DInSAR, to point out the relevance of a consequent analysis over an area of a NPP. Furthermore we present results of case studies from international projects (TERRAFIRMA) as well as preliminary results from the Krsko NPP in Slovenia. Finally, we provide a outlook into present and future trends concerning the use of freely

  19. INPRO Activities on Development of Advanced Tools to Support Judgment Aggregation for Comparative Evaluation of Nuclear Energy Systems

    Directory of Open Access Journals (Sweden)

    V. Kuznetsov

    2015-01-01

    Full Text Available This paper presents first results of the INPRO Collaborative Project on Key Indicators for Innovative Nuclear Energy Systems, which has the objective to develop guidance and tools for comparative evaluation of the status, prospects, benefits, and risks associated with development of innovative nuclear technologies for a more distant future. Presented results illustrate expedience of application of the multicriteria decision analysis methods, which are able to provide the added value to comparative assessment of nuclear energy systems. First, the paper presents a short review of the multicriteria decision analysis methods appropriate to support judgment aggregation within comparative evaluations of nuclear energy systems based on key indicators and highlights the methodology to perform such assessments. Second, a set of key indicators elaborated in the INPRO Collaborative Project on Global Architecture of Innovative Nuclear Energy Systems Based on Thermal and Fast Reactors Including a Closed Fuel Cycle (GAINS were evaluated for comparative evaluation of nuclear energy system evolution scenarios. Third, a numerical example is presented of application of the selected key indicators, methods, and tools for judgment aggregation in comparative assessment of the GAINS nuclear energy systems.

  20. Report on the workshop "Decay spectroscopy at CARIBU: advanced fuel cycle applications, nuclear structure and astrophysics". 14-16 April 2011, Argonne National Laboratory, USA.

    Energy Technology Data Exchange (ETDEWEB)

    Kondev, F.; Carpenter, M.P.; Chowdhury, P.; Clark, J.A.; Lister, C.J.; Nichols, A.L.; Swewryniak, D. (Nuclear Engineering Division); (Univ. of Massachusetts); (Univ. of Surrey)

    2011-10-06

    A workshop on 'Decay Spectroscopy at CARIBU: Advanced Fuel Cycle Applications, Nuclear Structure and Astrophysics' will be held at Argonne National Laboratory on April 14-16, 2011. The aim of the workshop is to discuss opportunities for decay studies at the Californium Rare Isotope Breeder Upgrade (CARIBU) of the ATLAS facility with emphasis on advanced fuel cycle (AFC) applications, nuclear structure and astrophysics research. The workshop will consist of review and contributed talks. Presentations by members of the local groups, outlining the status of relevant in-house projects and availabile equipment, will also be organized. time will also be set aside to discuss and develop working collaborations for future decay studies at CARIBU. Topics of interest include: (1) Decay data of relevance to AFC applications with emphasis on reactor decay heat; (2) Discrete high-resolution gamma-ray spectroscopy following radioactive decya and related topics; (3) Calorimetric studies of neutron-rich fission framgents using Total ABsorption Gamma-Ray Spectrometry (TAGS) technique; (4) Beta-delayed neutron emissions and related topics; and (5) Decay data needs for nuclear astrophysics.

  1. Interagency Advanced Power Group, Joint Electrical and Nuclear Working Group, meeting minutes, November 16--17, 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-12-31

    Reports on soldier power R&D review, N-MCT power electronic building blocks, silicon carbide power semiconductor work, and ground based radar were made to the Power Conditioning Panel. An introduction to high temperature electronics needs, research and development was made to the High Temperature Electronics Subcommittee. The Pulse Power Panel received reports on the navy ETC gun, and army pulse power. The Superconductivity Panel received reports on high-tc superconducting wires, superconducting magnetic energy storage, and superconducting applications. The Nuclear Working Group received presentations on the Topaz nuclear power program, and space nuclear work in the Department of Energy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-01

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

  4. Contributions of the National Institute of Nuclear Research to the advance of Science and Technology in Mexico. Commemorative edition 2010; Contribuciones del Instituto Nacional de Investigaciones Nucleares al avance de la ciencia y la tecnologia en Mexico. Edicion conmemorativa 2010

    Energy Technology Data Exchange (ETDEWEB)

    Duque M, G.; Jimenez R, M.; Monroy G, F.; Romero H, S.; Serment G, J. (ed.) [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-07-01

    From the second decade of the X X century the applications of the nuclear energy have been important part of the scientific and technological patrimony in Mexico. Records exist with regard to the use of the radioisotopes and the radiations in our country in that time, and in a formal way until the year of 1950, in a process that culminates with the creation of the Comision Nacional de Energia Nuclear (CNEN. January 1, 1956). In January 12, 1972 were published the Organic Law that created to the Instituto Nacional de Energia Nuclear, being responsible for the works that the CNEN developed. The current Instituto Nacional de Investigaciones Nucleares (ININ) was constituted starting from the Regulation Law of the constitutional Article 27 in nuclear matter of January 26, 1979, abrogated and substituted by the Law in force of February 4, 1985. In this lapse they were undertaken multitude of projects with results and diverse achievements. From their creation, the mission of the ININ and the previous institutions has been to realize research in science and nuclear technology, to promote their peaceful uses and to diffuse the achieved advances, always searching for to link them to the economic, social, scientific and technological development of the country. In this occasion with the purpose of participating in the commemoration of the bicentennial of the independence and centennial of the Mexican revolution in our country, the ININ decided to publish this work, directed to a wide public, with the intention of providing a vision the most complete and appropriate possible of the activities in research and technology that it is carries out at the moment. This work also seeks to be a diffusion instrument of the tasks that they are carried out in the institute, in diverse subjects as: the basic research, the nuclear applications in the health, the agriculture and the industry, the studies on the contamination and the environment; the dosimetry; the radiological protection; as

  5. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  6. Expression of receptor for advanced glycation endproducts and nuclear factor κB in brain hippocampus of rat with chronic fluorosis

    Institute of Scientific and Technical Information of China (English)

    张凯琳

    2014-01-01

    Objective To investigate the expressions of receptor for advanced glycation endproducts(RAGE)and nuclear factorκB(NF-κB)in brain hippocampus of rat with chronic fluorosis,and to reveal the mechanism of brain damage resulted from chronic fluorosis.Methods Sixty clean grade SD rats were randomly divided to three groups(20 rats in each group,10 female and 10 male)fed with different contents of fluoride,control group with normal tap-water(<0.5 mg/L fluoride),

  7. Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS) Code Verification and Validation Data Standards and Requirements: Fluid Dynamics Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Greg Weirs; Hyung Lee

    2011-09-01

    V&V and UQ are the primary means to assess the accuracy and reliability of M&S and, hence, to establish confidence in M&S. Though other industries are establishing standards and requirements for the performance of V&V and UQ, at present, the nuclear industry has not established such standards or requirements. However, the nuclear industry is beginning to recognize that such standards are needed and that the resources needed to support V&V and UQ will be very significant. In fact, no single organization has sufficient resources or expertise required to organize, conduct and maintain a comprehensive V&V and UQ program. What is needed is a systematic and standardized approach to establish and provide V&V and UQ resources at a national or even international level, with a consortium of partners from government, academia and industry. Specifically, what is needed is a structured and cost-effective knowledge base that collects, evaluates and stores verification and validation data, and shows how it can be used to perform V&V and UQ, leveraging collaboration and sharing of resources to support existing engineering and licensing procedures as well as science-based V&V and UQ processes. The Nuclear Energy Knowledge base for Advanced Modeling and Simulation (NE-KAMS) is being developed at the Idaho National Laboratory in conjunction with Bettis Laboratory, Sandia National Laboratories, Argonne National Laboratory, Utah State University and others with the objective of establishing a comprehensive and web-accessible knowledge base to provide V&V and UQ resources for M&S for nuclear reactor design, analysis and licensing. The knowledge base will serve as an important resource for technical exchange and collaboration that will enable credible and reliable computational models and simulations for application to nuclear power. NE-KAMS will serve as a valuable resource for the nuclear industry, academia, the national laboratories, the U.S. Nuclear Regulatory Commission (NRC) and

  8. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    Under contract with the U.S. Department of Energy, the College of Engineering at the University of South Carolina has conducted an assessment of educational needs among engineers working in nuclear fuel cycle related areas. The study was initiated as a regional effort focusing on the concentration of nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. The project was envisioned as providing base line information for the eventual implementation of a program in line with the needs of the Southeast's nuclear community. Specific objectives were to establish the content of such a program and to determine those specialized features which would make the program most attractive and useful.

  9. Emergency Preparedness technology support to the Health and Safety Executive (HSE), Nuclear Installations Inspectorate (NII) of the United Kingdom. Appendix A

    Energy Technology Data Exchange (ETDEWEB)

    O`Kula, K.R.

    1994-03-01

    The Nuclear Installations Inspectorate (NII) of the United Kingdom (UK) suggested the use of an accident progression logic model method developed by Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) for K Reactor to predict the magnitude and timing of radioactivity releases (the source term) based on an advanced logic model methodology. Predicted releases are output from the personal computer-based model in a level-of-confidence format. Additional technical discussions eventually led to a request from the NII to develop a proposal for assembling a similar technology to predict source terms for the UK`s advanced gas-cooled reactor (AGR) type. To respond to this request, WSRC is submitting a proposal to provide contractual assistance as specified in the Scope of Work. The work will produce, document, and transfer technology associated with a Decision-Oriented Source Term Estimator for Emergency Preparedness (DOSE-EP) for the NII to apply to AGRs in the United Kingdom. This document, Appendix A is a part of this proposal.

  10. Army Gas-Cooled Reactor Systems Program. ML-1 analytical design report. Volume II. Systems analysis: heat transfer and fluid flow

    Energy Technology Data Exchange (ETDEWEB)

    None

    1961-01-01

    The analysis preceding and supporting the design of the cooling system of the ML-1, a mobile, low-power, nuclear power plant, is described in sufficient detail for an engineer to follow the development of the design. Test results and similar data are used to support the calculations whenever possible.

  11. Benchmarking of thermal hydraulic loop models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES), phase-I: Isothermal steady state forced convection

    Science.gov (United States)

    Cho, Jae Hyun; Batta, A.; Casamassima, V.; Cheng, X.; Choi, Yong Joon; Hwang, Il Soon; Lim, Jun; Meloni, P.; Nitti, F. S.; Dedul, V.; Kuznetsov, V.; Komlev, O.; Jaeger, W.; Sedov, A.; Kim, Ji Hak; Puspitarini, D.

    2011-08-01

    As highly promising coolant for new generation nuclear reactors, liquid Lead-Bismuth Eutectic has been extensively worldwide investigated. With high expectation about this advanced coolant, a multi-national systematic study on LBE was proposed in 2007, which covers benchmarking of thermal hydraulic prediction models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES). This international collaboration has been organized by OECD/NEA, and nine organizations - ENEA, ERSE, GIDROPRESS, IAEA, IPPE, KIT/IKET, KIT/INR, NUTRECK, and RRC KI - contribute their efforts to LACANES benchmarking. To produce experimental data for LACANES benchmarking, thermal-hydraulic tests were conducted by using a 12-m tall LBE integral test facility, named as Heavy Eutectic liquid metal loop for integral test of Operability and Safety of PEACER (HELIOS) which has been constructed in 2005 at the Seoul National University in the Republic of Korea. LACANES benchmark campaigns consist of a forced convection (phase-I) and a natural circulation (phase-II). In the forced convection case, the predictions of pressure losses based on handbook correlations and that obtained by Computational Fluid Dynamics code simulation were compared with the measured data for various components of the HELIOS test facility. Based on comparative analyses of the predictions and the measured data, recommendations for the prediction methods of a pressure loss in LACANES were obtained. In this paper, results for the forced convection case (phase-I) of LACANES benchmarking are described.

  12. Advances in nuclear particle dosimetry for radiation protection and medicine - Ninth Symposium on Neutron Dosimetry (Editorial Material, English)

    Energy Technology Data Exchange (ETDEWEB)

    Zoetelief, J; Bos, A J.; Schuhmacher, H; McDonald, Joseph C.; Schultz, F W.; Pihet, P

    2004-12-15

    The Ninth Symposium on Neutron Dosimetry has been expanded to cover not only neutron radiation but heavy charged particle dosimetry as well. The applications are found in such fields as radiation protection, aircrew dosimetry, medicine, nuclear power and accelerator health physics. Scientists from many countries from around the world presented their work, and described the latest developments in techniques and instrumentation.

  13. Advanced Borobond™ Shields for Nuclear Materials Containment and Borobond™ Immobilization of Volatile Fission Products - Final CRADA Report

    Energy Technology Data Exchange (ETDEWEB)

    Wagh, Arun S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-05-19

    Borobond is a company-proprietary material developed by the CRADA partner in collaboration with Argonne, and is based on Argonne's Ceramicrete technology. It is being used by DOE for nuclear materials safe storage, and Boron Products, LLC is the manufacturer and supplier of Borobond.

  14. 77 FR 34194 - Advance Notification to Native American Tribes of Transportation of Certain Types of Nuclear Waste

    Science.gov (United States)

    2012-06-11

    ... required for access to Safeguards Information (SGI). The participating Tribal government is required to... this final rule. You may access information and comment submittals related to this final rule, which... access to a nuclear power facility or access to Safeguards Information,'' or Sec. 73.59, ``Relief...

  15. MCNP Super Lattice Method for VHTR ORIGEN2.2 Nuclear Library Improvement Based on ENDF/B-VII

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; J. R. Parry

    2010-10-01

    The advanced Very High Temperature gas-cooled Reactor (VHTR) achieves simplification of safety through reliance on innovative features and passive systems. One of the VHTRs innovative features is the reliance on ceramic-coated fuel particles to retain the fission products under extreme accident conditions. The effect of the random fuel kernel distribution in the fuel prismatic block creates a double-heterogeneous lattice, which needs to be addressed through the use of the newly developed prismatic super Kernel-by-Kernel Fuel (KbKF) lattice model method. Based on the new ENDF/B-VII nuclear cross section evaluated data, the developed KbKF super lattice model was then used with MCNP to calculate the material isotopes neutron reaction rates, such as, (n,?); (n,n’); (n,2n’); (n,f); (n,p); (n,?). Then, the MCNP-calculated results are rearranged to generate a set of new libraries “VHTRXS.lib,” for the ORIGEN2.2 isotopes depletion and build-up analysis code. The libraries contain one group cross section data for the structural light elements, actinides, and fission products that can be applied in the VHTR related fuel burnup and material transmutation analysis codes. The efficiency and ease of use of the MCNP method to generate and update the ORIGEN2.2 one-group spectrum weighed cross section library for VHTR was demonstrated.

  16. Applications of nuclear physics

    Science.gov (United States)

    Hayes, A. C.

    2017-02-01

    Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applications of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.

  17. Radiation Effects in Nuclear Ceramics

    Directory of Open Access Journals (Sweden)

    L. Thomé

    2012-01-01

    Full Text Available Due to outstanding physicochemical properties, ceramics are key engineering materials in many industrial domains. The evaluation of the damage created in ceramics employed in radiative media is a challenging problem for electronic, space, and nuclear industries. In this latter field, ceramics can be used as immobilization forms for radioactive wastes, inert fuel matrices for actinide transmutation, cladding materials for gas-cooled fission reactors, and structural components for fusion reactors. Information on the radiation stability of nuclear materials may be obtained by simulating the different types of interactions involved during the slowing down of energetic particles with ion beams delivered by various types of accelerators. This paper presents a review of the radiation effects occurring in nuclear ceramics, with an emphasis on recent results concerning the damage accumulation processes. Energetic ions in the KeV-GeV range are used to explore the nuclear collision (at low energy and electronic excitation (at high energy regimes. The recovery by electronic excitation of the damage created by ballistic collisions (SHIBIEC process is also addressed.

  18. Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores; Homogeneisation de modeles de transferts thermiques et radiatifs: application au coeur des reacteurs a caloporteur gaz

    Energy Technology Data Exchange (ETDEWEB)

    El Ganaoui, K

    2006-09-15

    In the context of homogenization theory we treat some heat transfer problems involving unusual (according to the homogenization) boundary conditions. These problems are defined in a solid periodic perforated domain where two scales (macroscopic and microscopic) are to be taken into account and describe heat transfer by conduction in the solid and by radiation on the wall of each hole. Two kinds of radiation are considered: radiation in an infinite medium (non-linear problem) and radiation in cavity with grey-diffuse walls (non-linear and non-local problem). The derived homogenized models are conduction problems with an effective conductivity which depend on the considered radiation. Thus we introduce a framework (homogenization and validation) based on mathematical justification using the two-scale convergence method and numerical validation by simulations using the computer code CAST3M. This study, performed for gas cooled reactors cores, can be extended to other perforated domains involving the considered heat transfer phenomena. (author)

  19. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  20. Study of advanced professional educational requirements relative to nuclear fuel cycle engineering in industry and government. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Jur, T.A.; Huhns, M.N.; Keating, D.A.; Orloff, D.I.; Rhodes, C.A.; Stanford, T.G.; Stephens, L.M.; Tatterson, G.B.; Van Brunt, V.

    1978-12-01

    An assessment was conducted of educational needs among engineers working in nuclear fuel cycle-related areas, focusing on the nuclear industry in the Southeast. Educational needs addressed were those at the post-baccalaureate professional level. As a result of the study, a list of subject areas has been compiled as best representing the current content of an educational program. In addition to identifying subject areas, a set of course descriptions and reference materials has been developed around each subject. Each course description contains information regarding objectives, anticipated audience, and prerequisites and offers a suggested course outline. An initial modest program of implementation is recommended which would continue to concentrate on the Southeast as a target area.

  1. A Pilot Study Investigating the Effects of Advanced Nuclear Power Plant Control Room Technologies: Methods and Qualitative Results

    Energy Technology Data Exchange (ETDEWEB)

    BLanc, Katya Le [Idaho National Lab. (INL), Idaho Falls, ID (United States); Powers, David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Joe, Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spielman, Zachary [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rice, Brandon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fitzgerald, Kirk [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Control room modernization is an important part of life extension for the existing light water reactor fleet. None of the 99 currently operating commercial nuclear power plants in the U.S. has completed a full-scale control room modernization to date. Nuclear power plant main control rooms for the existing commercial reactor fleet remain significantly analog, with only limited digital modernizations. Upgrades in the U.S. do not achieve the full potential of newer technologies that might otherwise enhance plant and operator performance. The goal of the control room upgrade benefits research is to identify previously overlooked benefits of modernization, identify candidate technologies that may facilitate such benefits, and demonstrate these technologies through human factors research. This report describes a pilot study to test upgrades to the Human Systems Simulation Laboratory at INL.

  2. Recent nuclear technology advances of GE-H and GNF in partnership with CFE's Laguna Verde 2

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G.; Iwamoto, T.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel - Americas LLC, 3901 Castle Hayne Road, Wilmington, 28401 North Carolina (United States); Calleros M, G., E-mail: Gabriel.Cuevas-Vivas@age.com [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Veracruz-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2015-09-15

    This paper is presented based on work stemming from the long-standing technical partnership between Comision Federal de Electricidad (CFE) and Global Nuclear Fuel - Americas LLC (GNF), and this paper's purpose is to provide valuable information for the nuclear industry. First, a one-to-one comparison of measurements from Gamma Thermometers (G Ts) installed in Laguna Verde 2 (LV-2) core against axially corresponding Traverse In-core Probe (Tip) measurements is remarkably close. This longest running G T plant validation for the Economic Simplified Boiling Water Reactor (ESBWR) power monitoring devices is also proving that the devices have the potential to simplify power measurements in Boiling Water Reactors (BWRs). Second, code validation results show remarkable reduction of radial power uncertainty with respect to current technology when new GNF codes for three-dimensional core simulations are compared to Tip power measurements. The LV-2 core is of paramount importance for BWR data analysis since the core has been loaded from the start of operation with different GNF fuel bundle types, fuel management has been designed with GNF codes and on-line core monitoring has been performed with GNF's 3-Dimensional Monicore system. This validation is also reliable since gamma Tips are used for local power measurements rather than thermal neutron Tips, which are more sensitive to local turbulence and positional changes. A nuclear/thermal-hydraulic core simulation of GNF fuel is undertaken from the beginning of operation of LV-2 (cycle 1) through the range of different power scenarios including low power, original licensed thermal power, power up rate and extended power up rate. Results confirm the conclusions from previous validations in different BWRs and support GNF efforts for licensing the new generation of nuclear engineering codes. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Simos, N.

    2011-05-01

    operating envelope of both fission and fusion reactors. In advanced fission reactors composite materials are being designed in an effort to extend the life and improve the reliability of fuel rod cladding as well as structural materials. Composites are being considered for use as core internals in the next generation of gas-cooled reactors. Further, next-generation plasma-fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER) will rely on the capabilities of advanced composites to safely withstand extremely high neutron fluxes while providing superior thermal shock resistance.

  4. Research on the Application of Advanced Construction Techniques for Nuclear Power Plants%核电厂先进建造技术的应用研究

    Institute of Scientific and Technical Information of China (English)

    刘志弢; 范霁红

    2013-01-01

    In this paper, the research is carried out on the advanced techniques for the construction of nuclear power plants. In combination of the actual engineering situation, the main characteristics and impacts of the techniques are analyzed such as the foundation pit excavation by blasting, the once-off complete concrete placement of raft foundation, the steel plate reinforced concrete structure, the open top construction, the advanced welding process, the modern engineering materials, as well as the prefabrication preassembly and modularization. The trends of the advanced construction techniques are also predicted.%研究了核电厂先进建造技术的现状;结合工程实际,分析了核电厂基坑负挖的爆破施工技术、筏基大体积混凝土一次性整体浇注、钢板混凝土结构、开顶法施工技术、先进焊接技术、现代建筑工程材料、预制造-预组装-模块化(PPM)技术等的主要特点和对工程建造的影响;归纳了核电厂先进建造技术发展趋势的主要特征.

  5. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    Science.gov (United States)

    Ioffe, B. L.; Kochurov, B. P.

    2012-02-01

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  6. Gordon Conference on Nuclear Research

    Energy Technology Data Exchange (ETDEWEB)

    Austin, S.M.

    1983-09-01

    Session topics were: quarks and nuclear physics; anomalons and anti-protons; the independent particle structure of nuclei; relativistic descriptions of nuclear structure and scattering; nuclear structure at high excitation; advances in nuclear astrophysics; properties of nuclear material; the earliest moments of the universe; and pions and spin excitations in nuclei.

  7. Role of nuclear grade graphite in controlling oxidation in modular HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Windes, Willaim [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kane, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of core environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.

  8. Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

    Energy Technology Data Exchange (ETDEWEB)

    Reynard-Carette, C. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 (France)

    2015-07-01

    Research and development on nuclear fuel behavior under irradiations and accelerated ageing of structure materials is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR) currently under construction in the South of France in the CEA Cadarache research centre will offer a real opportunity to perform R and D programs and hence will crucially contribute to the selection, optimization and qualification of innovative materials and fuels. To perform such programs advanced accurate and innovative experiments, irradiation devices that contain material and fuel samples are required to be set up inside or beside the reactor core. These experiments needs beforehand in situ and on line sophisticated measurements to accurately reach specific and determining parameters such as thermal and fast neutron fluxes, nuclear heating and temperature conditions to precisely monitor and control the conducted assays. Consequently, since 2009 CEA and Aix-Marseille University collaborate in order to design and develop a new multi-sensor device which will be dedicated to measuring profiles of such conditions inside the experimental channels of the JHR. These works are performed in the framework of two complementary joint research programs called MAHRI-BETHY and INCORE. These programs couple experimental studies carried out both out-of nuclear fluxes (in laboratory) and under irradiation conditions (in OSIRIS MTR reactor in France and MARIA MTR reactor in Poland) with numerical works realized by thermal simulations (CAST3M code) and Monte Carlo simulations (MCNP code). These programs deal with three main aims. The first one corresponds to the design and/or the test of new in-pile instrumentation. The second one concerns the development of advanced calibration procedures in particular in the case of one specific sensor: a differential calorimeter used to quantify

  9. Nuclear astrophysics

    CERN Document Server

    Arnould, M

    1999-01-01

    Nuclear astrophysics is that branch of astrophysics which helps understanding some of the many facets of the Universe through the knowledge of the microcosm of the atomic nucleus. In the last decades much advance has been made in nuclear astrophysics thanks to the sometimes spectacular progress in the modelling of the structure and evolution of the stars, in the quality and diversity of the astronomical observations, as well as in the experimental and theoretical understanding of the atomic nucleus and of its spontaneous or induced transformations. Developments in other sub-fields of physics and chemistry have also contributed to that advance. Many long-standing problems remain to be solved, however, and the theoretical understanding of a large variety of observational facts needs to be put on safer grounds. In addition, new questions are continuously emerging, and new facts endanger old ideas. This review shows that astrophysics has been, and still is, highly demanding to nuclear physics in both its experime...

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

    Institute of Scientific and Technical Information of China (English)

    陈闽烽; 张海波

    2014-01-01

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

  11. Impact of Nuclear Data Uncertainties on Advanced Fuel Cycles and their Irradiated Fuel - a Comparison between Libraries

    Science.gov (United States)

    Díez, C. J.; Cabellos, O.; Martínez, J. S.

    2014-04-01

    The uncertainties on the isotopic composition throughout the burnup due to the nuclear data uncertainties are analysed. The different sources of uncertainties: decay data, fission yield and cross sections; are propagated individually, and their effect assessed. Two applications are studied: EFIT (an ADS-like reactor) and ESFR (Sodium Fast Reactor). The impact of the uncertainties on cross sections provided by the EAF-2010, SCALE6.1 and COMMARA-2.0 libraries are compared. These Uncertainty Quantification (UQ) studies have been carried out with a Monte Carlo sampling approach implemented in the depletion/activation code ACAB. Such implementation has been improved to overcome depletion/activation problems with variations of the neutron spectrum.

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

    Science.gov (United States)

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

    2011-06-01

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

  13. Effects of a Mixed Zone on TGO Displacement Instabilities of Thermal Barrier Coatings at High Temperature in Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Jian Wang

    2016-01-01

    Full Text Available Thermally grown oxide (TGO, commonly pure α-Al2O3, formed on protective coatings acts as an insulation barrier shielding cooled reactors from high temperatures in nuclear energy systems. Mixed zone (MZ oxide often grows at the interface between the alumina layer and top coat in thermal barrier coatings (TBCs at high temperature dwell times accompanied by the formation of alumina. The newly formed MZ destroys interface integrity and significantly affects the displacement instabilities of TGO. In this work, a finite element model based on material property changes was constructed to investigate the effects of MZ on the displacement instabilities of TGO. MZ formation was simulated by gradually changing the metal material properties into MZ upon thermal cycling. Quantitative data show that MZ formation induces an enormous stress in TGO, resulting in a sharp change of displacement compared to the alumina layer. The displacement instability increases with an increase in the MZ growth rate, growth strain, and thickness. Thus, the formation of a MZ accelerates the failure of TBCs, which is in agreement with previous experimental observations. These results provide data for the understanding of TBC failure mechanisms associated with MZ formation and of how to prolong TBC working life.

  14. Nuclear astrophysics

    Energy Technology Data Exchange (ETDEWEB)

    Arnould, M. [Institut d' Astronomie et d' Astrophysique, Universite Libre de Bruxelles, Bruxelles (Belgium); Takahashi, K. [Max-Planck-Institut fuer Astrophysik, Garching (Germany)

    1999-03-01

    Nuclear astrophysics is that branch of astrophysics which helps understanding of the Universe, or at least some of its many faces, through the knowledge of the microcosm of the atomic nucleus. It attempts to find as many nuclear physics imprints as possible in the macrocosm, and to decipher what those messages are telling us about the varied constituent objects in the Universe at present and in the past. In the last decades much advance has been made in nuclear astrophysics thanks to the sometimes spectacular progress made in the modelling of the structure and evolution of the stars, in the quality and diversity of the astronomical observations, as well as in the experimental and theoretical understanding of the atomic nucleus and of its spontaneous or induced transformations. Developments in other subfields of physics and chemistry have also contributed to that advance. Notwithstanding the accomplishment, many long-standing problems remain to be solved, and the theoretical understanding of a large variety of observational facts needs to be put on safer grounds. In addition, new questions are continuously emerging, and new facts endangering old ideas. This review shows that astrophysics has been, and still is, highly demanding to nuclear physics in both its experimental and theoretical components. On top of the fact that large varieties of nuclei have to be dealt with, these nuclei are immersed in highly unusual environments which may have a significant impact on their static properties, the diversity of their transmutation modes, and on the probabilities of these modes. In order to have a chance of solving some of the problems nuclear astrophysics is facing, the astrophysicists and nuclear physicists are obviously bound to put their competence in common, and have sometimes to benefit from the help of other fields of physics, like particle physics, plasma physics or solid-state physics. Given the highly varied and complex aspects, we pick here some specific nuclear

  15. Nuclear astrophysics

    Science.gov (United States)

    Arnould, M.; Takahashi, K.

    1999-03-01

    Nuclear astrophysics is that branch of astrophysics which helps understanding of the Universe, or at least some of its many faces, through the knowledge of the microcosm of the atomic nucleus. It attempts to find as many nuclear physics imprints as possible in the macrocosm, and to decipher what those messages are telling us about the varied constituent objects in the Universe at present and in the past. In the last decades much advance has been made in nuclear astrophysics thanks to the sometimes spectacular progress made in the modelling of the structure and evolution of the stars, in the quality and diversity of the astronomical observations, as well as in the experimental and theoretical understanding of the atomic nucleus and of its spontaneous or induced transformations. Developments in other subfields of physics and chemistry have also contributed to that advance. Notwithstanding the accomplishment, many long-standing problems remain to be solved, and the theoretical understanding of a large variety of observational facts needs to be put on safer grounds. In addition, new questions are continuously emerging, and new facts endangering old ideas. This review shows that astrophysics has been, and still is, highly demanding to nuclear physics in both its experimental and theoretical components. On top of the fact that large varieties of nuclei have to be dealt with, these nuclei are immersed in highly unusual environments which may have a significant impact on their static properties, the diversity of their transmutation modes, and on the probabilities of these modes. In order to have a chance of solving some of the problems nuclear astrophysics is facing, the astrophysicists and nuclear physicists are obviously bound to put their competence in common, and have sometimes to benefit from the help of other fields of physics, like particle physics, plasma physics or solid-state physics. Given the highly varied and complex aspects, we pick here some specific nuclear

  16. Nuclear Power in China

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    China’s vigorous efforts to propel development of nuclear power are paying off as the country’s nuclear power sector advances at an amazing pace. At present, China has set up three enormous nuclear power bases, one each in Qinshan of Zhejiang Province, Dayawan of Guangdong

  17. Nuclear fact book

    Energy Technology Data Exchange (ETDEWEB)

    Hill, O. F.; Platt, A. M.; Robinson, J. V. [comps

    1983-05-01

    This reference provides significant highlights and summary facts in the following areas: general energy; nuclear energy; nuclear fuel cycle; uranium supply and enrichment; nuclear reactors; spent fuel and advanced repacking concepts; reprocessing; high-level waste; gaseous waste; transuranic waste; low-level waste; remedial action; transportation; disposal; radiation information; environment; legislation; socio-political aspects; conversion factors; and a glossary. (GHT)

  18. Design concepts and advanced telerobotics development for facilities in the back end of the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Feldman, M.J.

    1987-01-01

    In the Fuel Recycle Division at the Oak Ridge National Laboratory, a comprehensive remote systems development program has existed for the past seven years. The new remote technology under development is expected to significantly improve remote operations by extending the range of tasks accomplished by remote means and increasing the efficiency of remote work undertaken. Five areas of the development effort are primary contributors to the goal of higher operating efficiency for major facilities for the back end of the nuclear fuel cycle. These areas are the single-cell concept, the low-flow ventilation concept, television viewing, equipment-mounting racks, and force-reflecting manipulation. These somewhat innovative directions are products of a design process where the technical scenario to be accomplished, the remote equipment to accomplish the scenario, and the facility design to house the equipment, are considered in an iterative design process to optimize performance, maximize long-term costs effectiveness, and minimize initial capital outlay. 14 refs., 3 figs.

  19. Advances in high-resolution nuclear magnetic resonance methods in inhomogeneous magnetic fields using intermolecular multiple quantum coherences

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    Strong and extremely homogeneous static magnetic field is usually required for high-resolution nu-clear magnetic resonance (NMR). However, in the cases of in vivo and so on, the magnetic field inho-mogeneity owing to magnetic susceptibility variation in samples is unavoidable and hard to eliminate by conventional methods such as shimming. Recently, intermolecular multiple quantum coherences (iMQCs) have been employed to eliminate inhomogeneous broadening and obtain high-resolution NMR spectra, especially for in vivo samples. Compared to other high-resolution NMR methods, iMQC method exhibits its unique feature and advantage. It simultaneously holds information of chemical shifts, multiplet structures, coupling constants, and relative peak areas. All the information is often used to analyze and characterize molecular structures in conventional one-dimensional NMR spec-troscopy. In this work, recent technical developments including our results in this field are summarized; the high-resolution mechanism is analyzed and comparison with other methods based on interactions between spins is made; comments on the current situation and outlook on the research directions are also made.

  20. Application of complex engineering solutions through advanced composite innovation (for repair of degraded buried pipe at Vandellos II Nuclear Power Plant); Reparacion de tuberias de un sistema de servicios no esenciales con recubrimiento interno de fibra de carbono

    Energy Technology Data Exchange (ETDEWEB)

    Bueno, J. M.; Raji, B. B.

    2011-07-01

    This technical presentation is focused on introducing an engineering solution approach and identification of sensitivity of applications of advanced carbon fiber in a pressurized wet environment: Engineering design, quality assurance of installation, inspection, and a comprehensive testing program to validate and bench mark the design data and compliance with code requirements in nuclear power plants.

  1. Development of advanced direct perception displays for nuclear power plants to enhance monitoring, control and fault management

    Energy Technology Data Exchange (ETDEWEB)

    Jones, B.G.; Shaheen, S.; Moray, N. [and others

    1997-08-01

    Traditional Single-Sensor-Single Indicator (SSSI) displays are poorly matched to the cognitive abilities of operators, especially for large and complex systems. It is difficult for operators to monitor very large arrays of displays and controls, and to integrate the information displayed therein. In addition, standard operating procedures (SOPs) are bulky (running to many hundreds of pages) and difficult to use, and operators may become lost. For these reasons, and also because it is becoming increasingly difficult to find replacements for aging hardware components, there is a trend towards computerized graphical interfaces for nuclear power plants (NPPs). There is, however, little rational theory for display design in this domain. This report describes some recent theoretical developments and shows how to develop displays which will greatly reduce the cognitive load on the operator and allow the use of perceptual rather than cognitive mechanisms while using SON and to support state diagnosis and fault management. The report outlines the conceptual framework within which such a new approach could be developed, and provides an example of how the operating procedures for the start-up sequence of a NPP could be realized. A detailed description of a set of displays for a graphical interface for the SON of the feedwater system is provided as an example of how the proposed approach could be realized, and a general account of how it would fit into the overall start-up sequence is given. Examples of {open_quotes}direct perception{close_quotes} or {open_quotes}ecological{close_quotes} configural state space displays to support the use of the proposed direct manipulation SOP interface are provided, and also a critical discussion which identifies some difficulties which may be anticipated should the general approach herein advocated be adopted.

  2. The molecular gas content of the advanced S+E merger NGC 4441 - Evidence for an extended decoupled nuclear disc?

    CERN Document Server

    Juette, E; Hüttemeister, S

    2009-01-01

    Mergers between a spiral and an elliptical (S+E mergers) are poorly studied so far despite the importance for galaxy evolution. NGC4441 is a nearby candidate for an advanced remnant of such a merger, showing typical tidal structures like an optical tail and two shells as well as two HI tails. The study of the molecular gas content gives clues on the impact of the recent merger event on the star formation. Simulations of S+E mergers predict contradictory scenarios concerning the strength and the extent of an induced starburst. Thus, observations of the amount and the distribution of the molecular gas, the raw material of star formation, are needed to understand the influence of the merger on the star formation history. 12CO and 13CO (1-0) and (2-1) observations were obtained using the Onsala Space Observatory 20m and IRAM 30m telescope as well as the Plateau de Bure interferometer. These data allow us to carry out a basic analysis of the molecular gas properties such as estimates of the molecular gas mass, its...

  3. 间接氦气透平循环高温堆紧急停堆动态仿真%Dynamic simulation for scram of high temperature gas-cooled reactor with indirect helium turbine cycle system

    Institute of Scientific and Technical Information of China (English)

    李文龙; 解衡

    2011-01-01

    A dynamic analysis code for this system was developed after the mathematical modeling and programming of important equipment of 10 MW High Temperature Gas-cooled Reactor Helium Turbine Power Generation (HTR-10GT), such as reactor core, heat exchanger and turbine-compressor system. A scram accident caused by a 0. 1 $ reactivity injection at 5 second was simulated. The results show that the design emergency shutdown plan for this system is safe and reasonable and that the design of bypass valve has a large safety margin.%通过对10 MW高温气冷堆氦气透平发电装置(HTR-10GT)的堆芯、热交换器和透平压气机组等主要设备的数学建模和程序编制,初步建立起了一套模拟该装置瞬态特性的仿真程序.通过对该装置于5s时刻堆内引入0.1$阶跃正反应性引发的紧急停堆事故的瞬态模拟,初步验证了该装置紧急停堆预案设置的安全性和合理性,证明了旁路快开阀的设计预案具有较大的安全裕度.

  4. ADVANCED NUCLEAR FUEL CYCLE EFFECTS ON THE TREATMENT OF UNCERTAINTY IN THE LONG-TERM ASSESSMENT OF GEOLOGIC DISPOSAL SYSTEMS - EBS INPUT

    Energy Technology Data Exchange (ETDEWEB)

    Sutton, M; Blink, J A; Greenberg, H R; Sharma, M

    2012-04-25

    encapsulated in borosilicate glass. Because the heat load of the glass was much less than the PWR and BWR assemblies, the glass waste form was able to be co-disposed with the open cycle waste, by interspersing glass waste packages among the spent fuel assembly waste packages. In addition, the Yucca Mountain repository was designed to include some research reactor spent fuel and naval reactor spent fuel, within the envelope that was set using the commercial reactor assemblies as the design basis waste form. This milestone report supports Sandia National Laboratory milestone M2FT-12SN0814052, and is intended to be a chapter in that milestone report. The independent technical review of this LLNL milestone was performed at LLNL and is documented in the electronic Information Management (IM) system at LLNL. The objective of this work is to investigate what aspects of quantifying, characterizing, and representing the uncertainty associated with the engineered barrier are affected by implementing different advanced nuclear fuel cycles (e.g., partitioning and transmutation scenarios) together with corresponding designs and thermal constraints.

  5. Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model

    Energy Technology Data Exchange (ETDEWEB)

    Denia Djokic; Steven J. Piet; Layne F. Pincock; Nick R. Soelberg

    2013-02-01

    This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system , and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity.

  6. Advanced Nuclear Steam Turbine for Highest Power Output%先进的具高出力的核电汽轮机

    Institute of Scientific and Technical Information of China (English)

    Andreas Wichtmann; Norbert Schindler; Wilfried Ulm

    2003-01-01

    介绍了西门子设计制造先进大功率核电汽轮机的长期运行经验,并发展半速饱和蒸汽核电汽轮机,功率达1700MW,末级长叶片高度为1829mm,从设计、工艺、材料选择等方面论述了提高机组出力、提高机组效率、减少腐蚀的方法,其中包括套装叶轮设计、3DSTM叶片技术、高压透平和低压透平等.%The paper introduces long operational experience of nuclear power plant at Siemens which has been developing the half speed saturated steam turboset for high power output up to 1700MWel with the longest blade length up to 1829mm/72 inch. The advanced design methods, process and appropriate material selection have been proved to increase the highest power output and efficiency, decrease erosion and corrosion. The features of the shrunk-on disk design, 3DSTM blade technique, high pressure turbine and low pressure turbine are also described in the paper.

  7. Regulation of alternative splicing of the receptor for advanced glycation endproducts (RAGE) through G-rich cis-elements and heterogenous nuclear ribonucleoprotein H.

    Science.gov (United States)

    Ohe, Kazuyo; Watanabe, Takuo; Harada, Shin-ichi; Munesue, Seiichi; Yamamoto, Yasuhiko; Yonekura, Hideto; Yamamoto, Hiroshi

    2010-05-01

    Receptor for advanced glycation endproducts (RAGE) is a cell-surface receptor. The binding of ligands to membrane-bound RAGE (mRAGE) evokes cellular responses involved in various pathological processes. Previously, we identified a novel soluble form, endogenous secretory RAGE (esRAGE) generated by alternative 5' splice site selection in intron 9 that leads to extension of exon 9 (exon 9B). Because esRAGE works as an antagonistic decoy receptor, the elucidation of regulatory mechanism of the alternative splicing is important to understand RAGE-related pathological processes. Here, we identified G-rich cis-elements within exon 9B for regulation of the alternative splicing using a RAGE minigene. Mutagenesis of the G-rich cis-elements caused a drastic increase in the esRAGE/mRAGE ratio in the minigene-transfected cells and in loss of binding of the RNA motif to heterogenous nuclear ribonucleoprotein (hnRNP) H. On the other hand, the artificial introduction of a G-stretch in exon 9B caused a drastic decrease in the esRAGE/mRAGE ratio accompanied by the binding of hnRNP H to the RNA motif. Thus, the G-stretches within exon 9B regulate RAGE alternative splicing via interaction with hnRNP H. The findings should provide a molecular basis for the development of medicines for RAGE-related disorders that could modulate esRAGE/mRAGE ratio.

  8. Study and Development of a Simulation System for Dynamic Evaluation on Man-machine Interface Design of Advanced Main Control Rooms of Nuclear Power Plants

    Institute of Scientific and Technical Information of China (English)

    YangXiaojing; ZhouZhiwei; ChenXiaoming; MaYuanle; LiFu; DongYujie; WuWei; OhiTadashi

    2005-01-01

    Since the man-machine interfaces (MMI) of a main control room provide the control platform of a nuclear power plant (NPP),the development of the design quality of MMIs plays a very important role in the operation of a NPP. With the development of digital technology, the development of the advanced main control rooms (AMCRs) has become an inexorable trend. Therefore, the positive and the negative effects of AMCRs on human factors engineering need to be evaluated. For this p~, a simulation system has been studied and developed to quantitatively evaluate a MMI design from the viewpoint of human factors. The simulation system takes advantage of computer simulation technology to simulate an operating process of an interaction between operators and a MMI design under an instruction of an operation procedure of the AMCR of a NPP. Meanwhile, the necessary data are recorded for evaluation. It integrates two editors and one simulator. In the paper, the simulation system is presented in detail. Furthermore, one sample is given to show the results of each of these three subsystems.

  9. Technical Overview on Main Helium Ventilator of High-Temperature Gas-Cooled Reactor Nuclear Power Plant%高温气冷堆核电站示范工程主氦风机技术概述

    Institute of Scientific and Technical Information of China (English)

    邹德宝

    2011-01-01

    介绍了当今世界核电技术的发展趋势及应用,提出了第四代核电技术需要满足的条件,并说明高温气冷堆是目前最有希望成为第四代先进核能系统的一种堆型;介绍了高温气冷堆的核心设备主氦风机的性能参数、结构、功能和技术先进性,并说明主氦风机研制的重要意义.

  10. Operation Characteristics Study of Modular High Temperature Gas-Cooled Reactor Nuclear Power Plant%模块式高温气冷堆核电站的运行特性研究

    Institute of Scientific and Technical Information of China (English)

    李海鹏; 黄晓津; 张良驹

    2009-01-01

    球床模块式高温气冷堆核电站(HTR-PM)由于具有多个模块,运行特性比单堆电站更复杂.利用集总参数方法建立了HTR-PM的动态模型,并利用该模型对电站的运行过程进行了仿真.升功率运行的仿真结果表明,蒸汽温度严重偏离了正常允许值.设计了1个基于蒸汽温度的简单控制器,仿真结果表明,该控制器能很好地对电站进行运行控制,结果令人满意.

  11. 某高温气冷堆核电厂结构地震反应分析%Seismic Response Analysis of High Temperature Gas Cooled Reactor Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    贺秋梅; 李小军; 张江伟; 李亚琦

    2014-01-01

    以某高温气冷堆核电厂结构为原型,利用有限元软件建立三维结构实体模型,开展了模态分析和弹性动力时程分析的数值计算,以探讨某高温气冷堆核电厂的结构特性和抗震性能,并重点分析了在三向地震动作用下高温气冷堆核电厂的加速度、位移反应时程和楼层反应谱.总体上看,高温气冷堆核电厂在两个水平向的刚度比较均匀,楼层反应接近,布局较为合理;在三向地震动作用下,顶层中心点的竖向楼层反应均明显大于两水平方向楼层反应,因此在高温气冷堆核电厂结构设计中,应关注竖向地震动对核电厂地震反应带来的不利影响.

  12. Study on Related Problems of Nuclear Graphite for High Temperature Gas-Cooled Reactor%高温气冷堆核级石墨相关问题研究

    Institute of Scientific and Technical Information of China (English)

    赵木

    2014-01-01

    本文通过对石墨在高温气冷堆中的运行环境进行了分析,研究了在石墨堆内构件设计中的关键问题和在高温气冷堆单个模块及其未来发展中核级石墨的需求.从原料、成型及中子辐照等角度分析了核级石墨国产化研究方向.根据核级石墨目前的研发形势,进行了风险问题分析.

  13. HTGR nuclear heat source component design and experience

    Energy Technology Data Exchange (ETDEWEB)

    Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

    1982-05-01

    The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included.

  14. Preliminary Study on Insulating Design of Electrical Device in Helium for High Temperature Gas-Cooled Reactor%高温气冷堆氦气环境中电气设备绝缘设计研究

    Institute of Scientific and Technical Information of China (English)

    于晓丽; 杨小勇; 周世新; 王捷

    2011-01-01

    应用巴申定律研究了氦气的电气击穿特性,并与空气的绝缘特性进行比较.以高温气冷堆氦气透平发电系统电机腔室的设计参数为例,结合氦气的巴申曲线,对氦气条件下气体压力和极间距离的关系进行深入探讨,并提出氦气环境中电气设备绝缘设计需关注的问题.研究结果表明,氦气最小击穿电压为150~200 V,绝缘特性较差,电气设备绝缘结构设计应考虑氦气环境压力的影响,现有针对压水堆电站电气设备绝缘结构的验收准则和试验方法并不完全适用于氦气环境.%The breakdown performance of helium was studied by Paschen law, comparing with air. Combined with the operation parameter of generator in gas turbine coupled with high temperature gas-cooled reactor and the Paschen curve of helium, the relationship between pressure and insulating structure was discussed. The key points for the insulation design of the electrical device in helium were presented. The results show that the insulation performance for helium which lowest breakdown potentials is 150-200 V is much worse than that of air. The existing test and inspect guidelines of the insulation structure for the pressure water reactor can't be used for the helium. High permeability for helium may be an important reason to destroy the insulation structure.

  15. 高温气冷堆球床等效导热系数实验研究进展%Progress in Pebble Bed Heat Transfer Test Facility of High Temperature Gas-cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    任成; 杨星团; 李聪新; 孙艳飞

    2013-01-01

    Pebble bed effective thermal conductivity of the reactor core is a vital parameter which directly affects the maximum temperature of fuel and the temperature distribution of reactor core in high temperature gas-cooled reactor and plays a dominant role in afterheat removal. The experimental research of pebble bed effective thermal conductivity is helpful for improving the reactor analytical program, increasing the power of single reactor and the safe analysis of projects. The experiments and results of pebble bed effective thermal conductivity measurement at home and abroad are reviewed. The develop direction is also discussed.%堆芯球床等效导热系数是直接影响高温气冷堆燃料最高温度和堆芯温度分布的关键参数;在余热导出过程中起主导作用.开展球床等效导热系数的实验研究对于反应堆分析程序的完善、研究提高高温气冷堆单堆功率的可能性、以及工程的安全分析具有重要的意义.综述了国内外球床等效导热系数测量的研究现状,给出了清华大学HTR-PM三维堆芯球床等效导热系数测量实验最新成果,总结了各国实验的研究手段,对研究方向进行了讨论.

  16. Argon purge gas cooled by chill box

    Science.gov (United States)

    Spiro, L. W.

    1966-01-01

    Cooling argon purge gas by routing it through a shop-fabricated chill box reduces charring of tungsten inert gas torch head components. The argon gas is in a cooled state as it enters the torch and prevents buildup of char caused by the high concentrations of heat in the weld area during welding operations.

  17. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design.

  18. The reactor ALLEGRO and the sustainable nuclear energy in Central Europe

    Directory of Open Access Journals (Sweden)

    Gadó János

    2014-01-01

    Full Text Available The Visegrád-4 countries (CZ, HU, PL and SK would like to use nuclear energy on the long run. The construction of new Generation 3+ nuclear units probably belong in each country to this realm. These reactors will provide safe and cheap electric energy approximately until the end of the 21st century. In order to use nuclear energy in the 22nd century, sustainability of fuel supply shall be achieved by applying Generation 4 breeder reactors with fast spectrum. The corresponding research and development is organized now in the framework of the V4G4 Centre of Excellence establshed by the nuclear research institutes of the region with a strong technical support from the French CEA. The most important milestone of these efforts is the start-up of the ALLEGRO reactor that shall demonstrate the viability of the gas cooled fast reactor technology.

  19. A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-01

    United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

  20. Characterization of oil shale, isolated kerogen, and post-pyrolysis residues using advanced 13 solid-state nuclear magnetic resonance spectroscopy

    Science.gov (United States)

    Cao, Xiaoyan; Birdwell, Justin E.; Chappell, Mark A.; Li, Yuan; Pignatello, Joseph J.; Mao, Jingdong

    2013-01-01

    Characterization of oil shale kerogen and organic residues remaining in postpyrolysis spent shale is critical to the understanding of the oil generation process and approaches to dealing with issues related to spent shale. The chemical structure of organic matter in raw oil shale and spent shale samples was examined in this study using advanced solid-state 13C nuclear magnetic resonance (NMR) spectroscopy. Oil shale was collected from Mahogany zone outcrops in the Piceance Basin. Five samples were analyzed: (1) raw oil shale, (2) isolated kerogen, (3) oil shale extract