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Sample records for advanced nuclear thermal

  1. Advanced Filter Technology For Nuclear Thermal Propulsion

    Science.gov (United States)

    Castillon, Erick

    2015-01-01

    The Scrubber System focuses on using HEPA filters and carbon filtration to purify the exhaust of a Nuclear Thermal Propulsion engine of its aerosols and radioactive particles; however, new technology may lend itself to alternate filtration options, which may lead to reduction in cost while at the same time have the same filtering, if not greater, filtering capabilities, as its predecessors. Extensive research on various types of filtration methods was conducted with only four showing real promise: ionization, cyclonic separation, classic filtration, and host molecules. With the four methods defined, more research was needed to find the devices suitable for each method. Each filtration option was matched with a device: cyclonic separators for the method of the same name, electrostatic separators for ionization, HEGA filters, and carcerands for the host molecule method. Through many hours of research, the best alternative for aerosol filtration was determined to be the electrostatic precipitator because of its high durability against flow rate and its ability to cleanse up to 99.99% of contaminants as small as 0.001 micron. Carcerands, which are the only alternative to filtering radioactive particles, were found to be non-existent commercially because of their status as a "work in progress" at research institutions. Nevertheless, the conclusions after the research were that HEPA filters is recommended as the best option for filtering aerosols and carbon filtration is best for filtering radioactive particles.

  2. IMPULSE---an advanced, high performance nuclear thermal propulsion system

    International Nuclear Information System (INIS)

    Petrosky, L.J.; Disney, R.K.; Mangus, J.D.; Gunn, S.A.; Zweig, H.R.

    1993-01-01

    IMPULSE is an advanced nuclear propulsion engine for future space missions based on a novel conical fuel. Fuel assemblies are formed by stacking a series of truncated (U, Zr)C cones with non-fueled lips. Hydrogen flows radially inward between the cones to a central plenum connected to a high performance bell nozzle. The reference IMPULSE engine rated at 75,000 lb thrust and 1800 MWt weighs 1360 kg and is 3.65 meters in height and 81 cm in diameter. Specific impulse is estimated to be 1000 for a 15 minute life at full power. If longer life times are required, the operating temperature can be reduced with a concomitant decrease in specific impulse. Advantages of this concept include: well defined coolant paths without outlet flow restrictions; redundant orificing; very low thermal gradients and hence, thermal stresses, across the fuel elements; and reduced thermal stresses because of the truncated conical shape of the fuel elements

  3. Advanced modelling and numerical strategies in nuclear thermal-hydraulics

    International Nuclear Information System (INIS)

    Staedtke, H.

    2001-01-01

    The first part of the lecture gives a brief review of the current status of nuclear thermal hydraulics as it forms the basis of established system codes like TRAC, RELAP5, CATHARE or ATHLET. Specific emphasis is given to the capabilities and limitations of the underlying physical modelling and numerical solution strategies with regard to the description of complex transient two-phase flow and heat transfer conditions as expected to occur in PWR reactors during off-normal and accident conditions. The second part of the lecture focuses on new challenges and future needs in nuclear thermal-hydraulics which might arise with regard to re-licensing of old plants using bestestimate methodologies or the design and safety analysis of Advanced Light Water Reactors relying largely on passive safety systems. In order to meet these new requirements various advanced modelling and numerical techniques will be discussed including extended wellposed (hyperbolic) two-fluid models, explicit modelling of interfacial area transport or higher order numerical schemes allowing a high resolution of local multi-dimensional flow processes.(author)

  4. To MARS and Beyond with Nuclear Power - Design Concept of Korea Advanced Nuclear Thermal Engine Rocket

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Chang, Soon Heung

    2013-01-01

    The President Park of ROK has also expressed support for space program promotion, praising the success of NARO as evidence of a positive outlook. These events hint a strong signal that ROK's space program will be accelerated by the national eager desire. In this national eager desire for space program, the policymakers and the aerospace engineers need to pay attention to the advanced nuclear technology of ROK that is set to a major world nuclear energy country, even exporting the technology. The space nuclear application is a very much attractive option because its energy density is the most enormous among available energy sources in space. This paper presents the design concept of Korea Advanced Nuclear Thermal Engine Rocket (KANuTER) that is one of the advanced nuclear thermal rocket engine developing in Korea Advanced Institute of Science and Technology (KAIST) for space application. Solar system exploration relying on CRs suffers from long trip time and high cost. In this regard, nuclear propulsion is a very attractive option for that because of higher performance and already demonstrated technology. Although ROK was a late entrant into elite global space club, its prospect as a space racer is very bright because of the national eager desire and its advanced technology. Especially it is greatly meaningful that ROK has potential capability to launch its nuclear technology into space as a global nuclear energy leader and a soaring space adventurer. In this regard, KANuTER will be a kind of bridgehead for Korean space nuclear application

  5. To MARS and Beyond with Nuclear Power - Design Concept of Korea Advanced Nuclear Thermal Engine Rocket

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-05-15

    The President Park of ROK has also expressed support for space program promotion, praising the success of NARO as evidence of a positive outlook. These events hint a strong signal that ROK's space program will be accelerated by the national eager desire. In this national eager desire for space program, the policymakers and the aerospace engineers need to pay attention to the advanced nuclear technology of ROK that is set to a major world nuclear energy country, even exporting the technology. The space nuclear application is a very much attractive option because its energy density is the most enormous among available energy sources in space. This paper presents the design concept of Korea Advanced Nuclear Thermal Engine Rocket (KANuTER) that is one of the advanced nuclear thermal rocket engine developing in Korea Advanced Institute of Science and Technology (KAIST) for space application. Solar system exploration relying on CRs suffers from long trip time and high cost. In this regard, nuclear propulsion is a very attractive option for that because of higher performance and already demonstrated technology. Although ROK was a late entrant into elite global space club, its prospect as a space racer is very bright because of the national eager desire and its advanced technology. Especially it is greatly meaningful that ROK has potential capability to launch its nuclear technology into space as a global nuclear energy leader and a soaring space adventurer. In this regard, KANuTER will be a kind of bridgehead for Korean space nuclear application.

  6. NEPTUNE: A new software platform for advanced nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Guelfi, A.; Boucker, M.; Herard, J.M.; Peturaud, P.; Bestion, D.; Boudier, P.; Hervieu, E.; Fillion, P.; Grandotto, M.

    2007-01-01

    The NEPTUNE project constitutes the thermal-hydraulic part of the long-term Electricite de France and Commissariat a l'Energie Atomique joint research and development program for the next generation of nuclear reactor simulation tools. This program is also financially supported by the Institut de Radioprotection et Surete Nucleaire and AREVA NP. The project aims at developing a new software platform for advanced two-phase flow thermal hydraulics covering the whole range of modeling scales and allowing easy multi-scale and multidisciplinary calculations. NEPTUNE is a fully integrated project that covers the following fields: software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques, and performance of new experimental programs. The analysis of the industrial needs points out that three main simulation scales are involved. The system scale is dedicated to the overall description of the reactor. The component or subchannel scale allows three-dimensional computations of the main components of the reactors: cores, steam generators, condensers, and heat exchangers. The current generation of system and component codes has reached a very high level of maturity for industrial applications. The third scale, computational fluid dynamics (CFD) in open medium, allows one to go beyond the limits of the component scale for a finer description of the flows. This scale opens promising perspectives for industrial simulations, and the development and validation of the NEPTUNE CFD module have been a priority since the beginning of the project. It is based on advanced physical models (two-fluid or multi field model combined with interfacial area transport and two-phase turbulence) and modern numerical methods (fully unstructured finite volume solvers). For the system and component scales, prototype developments have also started, including new physical models and numerical methods. In addition to scale

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

    International Nuclear Information System (INIS)

    Song, C. H.; Baek, W. P.; Chung, M. K.

    2007-06-01

    The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. 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. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

  8. Advanced Fuel Forms for Nuclear Thermal Propulsion Applications

    Data.gov (United States)

    National Aeronautics and Space Administration — In NASA's 2014 Strategic Plan, Objective 1 specified the necessity to "expand human presence into the solar system and to the surface of mars to advance exploration,...

  9. 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.

  10. Development of CFD software for the simulation of thermal hydraulics in advanced nuclear reactors. Final report

    International Nuclear Information System (INIS)

    Bachar, Abdelaziz; Haslinger, Wolfgang; Scheuerer, Georg; Theodoridis, Georgios

    2015-01-01

    The objectives of the project were: Improvement of the simulation accuracy for nuclear reactor thermo-hydraulics by coupling system codes with three-dimensional CFD software; Extension of CFD software to predict thermo-hydraulics in advanced reactor concepts; Validation of the CFD software by simulation different UPTF TRAM-C test cases and development of best practice guidelines. The CFD module was based on the ANSYS CFD software and the system code ATHLET of GRS. All three objectives were met: The coupled ATHLET-ANSYS CFD software is in use at GRS and TU Muenchen. Besides the test cases described in the report, it has been used for other applications, for instance the TALL-3D experiment of KTH Stockholm. The CFD software was extended with material properties for liquid metals, and validated using existing data. Several new concepts were tested when applying the CFD software to the UPTF test cases: Simulations with Conjugate Heat Transfer (CHT) were performed for the first time. This led to better agreement between predictions and data and reduced uncertainties when applying temperature boundary conditions. The meshes for the CHT simulation were also used for a coupled fluid-structure-thermal analysis which was another novelty. The results of the multi-physics analysis showed plausible results for the mechanical and thermal stresses. The workflow developed as part of the current project can be directly used for industrial nuclear reactor simulations. Finally, simulations for two-phase flows with and without interfacial mass transfer were performed. These showed good agreement with data. However, a persisting problem for the simulation of multi-phase flows are the long simulation times which make use for industrial applications difficult.

  11. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Catana, Alexandru [RAAN, Institute for Nuclear Research, Str. Campului Nr. 1, Pitesti, Arges (Romania); Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel [University POLITEHNICA of Bucharest (Romania)

    2008-07-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D{sub 2}O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D{sub 2}O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

  12. Thermal-hydraulics analysis for advanced fuel to be used in Candu 600 nuclear reactors

    International Nuclear Information System (INIS)

    Catana, Alexandru; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2008-01-01

    Two Candu 600 pressure tube nuclear reactors cover about 17% of Romania's electricity demand. These nuclear reactors are moderated/cooled with D 2 O, fuelled on-power with Natural Uranium (NU) dioxide encapsulated in a standard (STD37) fuel bundle. High neutron economy is achieved using D 2 O as moderator and coolant in separated systems. To reduce fuel cycle costs, programs were initiated in Canada, S.Korea, Argentina and Romania for the design and build new fuel bundles able to accommodate different fuel compositions. Candu core structure and modular fuel bundles, permits flexible fuel cycles. The main expected achievements are: reduced fuel cycle costs, increased discharge burn-up, plutonium and minor actinides management, thorium cycle, use of recycled PWR and in the same time waste minimization and operating cost reduction. These new fuel bundles are to be used in already operated Candu reactors. Advanced fuel bundle were proposed: CANFLEX bundle (Canada, S-Korea); the Romanian 'SEU43' bundle (Fig 1). In this paper thermal-hydraulic analysis in sub-channel approach is presented for SEU43. Comparisons with standard (STD37) fuel bundles are made using SEU-NU for NU fuel composition and SEU-0.96, for recycled uranium (RU) fuel with 0.96% U-235. Extended and comprehensive analysis must be made in order to assess the TH behaviour of SEU43. In this paper, considering STD37, SEU43-NU and SEU43-0.96 fuel bundles, main TH parameters were analysed: pressure drop, fuel highest temperatures, coolant density, critical heat flux. Differences between these fuel types are outlined. Benefits are: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power. Safety margins must be, at least, conserved. (authors)

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

    International Nuclear Information System (INIS)

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

    2005-04-01

    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

  14. 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.

  15. Looking back on safety management in construction of advanced thermal nuclear power plant 'Fugen'

    International Nuclear Information System (INIS)

    Emori, Kengo

    1979-01-01

    The safety management of the advanced thermal nuclear power plant ''Fugen'' during the period from the preparation of plant construction in October, 1970, to the full power operation in March, 1979, is looked back and explained. Any large human and material accidents did not occur during the long construction time. The total numbers of persons and hours were 1.397 x 10 6 workers and 11.55 x 10 6 hours, respectively. The number of labor accidents was twenty with no dead person, the number of loss days was 645 days, the number of accident rate was 1.73, the intensity rate was 0.06, and the mean rate of labor accidents per year per 1000 workers was 6.4. The radiation exposure dose was 65.27 man-rem for the managed 1804 workers in total, during the testing and operating periods. These data show that the safety management for ''Fugen'' is very excellent, considering the following special features: 1) there were many works which were carried out for the first time, 2) the construction of the plant was conducted by five contractors taking partial charge, there were many kinds of construction works, the construction and testing periods were long, and the workers had to go to the site from Tsuruga city by car and bus. The organization of preventing disasters, the concrete implementation items for safety management, including the planning of activities, various meetings, patrol, education and training, the honoring system, the prevention of traffic accident and so on, and the results of actual safety management are explained with the reflection. (Nakai, Y.)

  16. Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

    Science.gov (United States)

    Magnusson, Per; Chen, Jiachao; Hoffelner, Wolfgang

    2009-12-01

    Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10 × 2 × 0.2 mm3 were investigated in a temperature range of 300 °C to 500 °C under irradiation, and significant creep strains were detected. At temperatures above 500 °C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900 °C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900 °C, the expected 104 hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloys.

  17. Recent advances in modeling and validation of nuclear thermal-hydraulics applications with NEPTUNE CFD - 15471

    International Nuclear Information System (INIS)

    Guingo, M.; Baudry, C.; Hassanaly, M.; Lavieville, J.; Mechitouna, N.; Merigoux, N.; Mimouni, S.; Bestion, D.; Coste, P.; Morel, C.

    2015-01-01

    NEPTUNE CFD is a Computational Multi-(Fluid) Dynamics code dedicated to the simulation of multiphase flows, primarily targeting nuclear thermo-hydraulics applications, such as the departure from nuclear boiling (DNB) or the two-phase Pressurized Thermal Shock (PTS). It is co-developed within the joint research/development project NEPTUNE (AREVA, CEA, EDF, IRSN) since 2001. Over the years, to address the aforementioned applications, dedicated physical models and numerical methods have been developed and implemented in the code, including specific sets of models for turbulent boiling flows and two-phase non-adiabatic stratified flows. This paper aims at summarizing the current main modeling capabilities of the code, and gives an overview of the associated validation database. A brief summary of emerging applications of the code, such as containment simulation during a potential severe accident or in-vessel retention, is also provided. (authors)

  18. 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

  19. Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    Nclear rockets improve the propellant efficiency more than twice compared to CRs and thus significantly reduce the propellant requirement. The superior efficiency of nuclear rockets is due to the combination of the huge energy density and a single low molecular weight propellant utilization. Nuclear Thermal Rockets (NTRs) are particularly suitable for manned missions to Mars because it satisfies a relatively high thrust as well as a high propellant efficiency. NTRs use thermal energy released from a nuclear fission reactor to heat a single low molecular weight propellant, i. e., Hydrogen (H{sub 2}) and then exhausted the extremely heated propellant through a thermodynamic nozzle to produce thrust. A propellant efficiency parameter of rocket engines is specific impulse (I{sub sp}) which represents the ratio of the thrust over the rate of propellant consumption. The difference of I{sub sp} makes over three times propellant savings of NTRs for a manned Mars mission compared to CRs. NTRs can also be configured to operate bimodally by converting the surplus nuclear energy to auxiliary electric power required for the operation of a spacecraft. Moreover, the concept and technology of NTRs are very simple, already proven, and safe. Thus, NTRs can be applied to various space missions such as solar system exploration, International Space Station (ISS) transport support, Near Earth Objects (NEOs) interception, etc. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The ROK has also begun the research for space nuclear systems as a volunteer of the international space race and a major world nuclear energy country. KANUTER is one of the advanced NTR engines currently under development at KAIST. This bimodal engine is operated in two modes of propulsion with 100 MW

  20. Consideration of a design optimization method for advanced nuclear power plant thermal-hydraulic components

    International Nuclear Information System (INIS)

    Ridluan, Artit; Tokuhiro, Akira; Manic, Milos; Patterson, Michael; Danchus, William

    2009-01-01

    In order to meet the global energy demand and also mitigate climate change, we anticipate a significant resurgence of nuclear power in the next 50 years. Globally, Generation III plants (ABWR) have been built; Gen' III+ plants (EPR, AP1000 others) are anticipated in the near term. The U.S. DOE and Japan are respectively pursuing the NGNP and MSFR. There is renewed interest in closing the fuel cycle and gradually introducing the fast reactor into the LWR-dominated global fleet. In order to meet Generation IV criteria, i.e. thermal efficiency, inherent safety, proliferation resistance and economic competitiveness, plant and energy conversion system engineering design have to increasingly meet strict design criteria with reduced margin for reliable safety and uncertainties. Here, we considered a design optimization approach using an anticipated NGNP thermal system component as a Case Study. A systematic, efficient methodology is needed to reduce time consuming trial-and-error and computationally-intensive analyses. We thus developed a design optimization method linking three elements; that is, benchmarked CFD used as a 'design tool', artificial neural networks (ANN) to accommodate non-linear system behavior and enhancement of the 'design space', and finally, response surface methodology (RSM) to optimize the design solution with targeted constraints. The paper presents the methodology including guiding principles, an integration of CFD into design theory and practice, consideration of system non-linearities (such as fluctuating operating conditions) and systematic enhancement of the design space via application of ANN, and a stochastic optimization approach (RSM) with targeted constraints. Results from a Case Study optimizing the printed circuit heat exchanger for the NGNP energy conversion system will be presented. (author)

  1. Advanced thermal management materials

    CERN Document Server

    Jiang, Guosheng; Kuang, Ken

    2012-01-01

    ""Advanced Thermal Management Materials"" provides a comprehensive and hands-on treatise on the importance of thermal packaging in high performance systems. These systems, ranging from active electronically-scanned radar arrays to web servers, require components that can dissipate heat efficiently. This requires materials capable of dissipating heat and maintaining compatibility with the packaging and dye. Its coverage includes all aspects of thermal management materials, both traditional and non-traditional, with an emphasis on metal based materials. An in-depth discussion of properties and m

  2. A review of modern advances in analyses and applications of single-phase natural circulation loop in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Basu, Dipankar N.; Bhattacharyya, Souvik; Das, P.K.

    2014-01-01

    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

  3. 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

  4. A new multi-scale platform for advanced nuclear thermal-hydraulics status and prospects of the Neptune project

    International Nuclear Information System (INIS)

    Bestion, D.; Boudier, P.; Hervieu, E.; Boucker, M.; Peturaud, P.; Guelfi, A.; Fillion, P.; Grandotto, M.; Herard, J.M.

    2005-01-01

    Full text of publication follows: Further to a thorough analysis of the industrial needs and of the limitations of current simulation tools, EDF (Electricite de France) and CEA (Commissariat a l'Energie Atomique) launched in 2001 a new long-term joint development program for the next generation of nuclear reactors simulation tools. The NEPTUNE Project, which constitutes the Thermal-Hydraulics part of this comprehensive program, aims at building a new software platform for advanced two-phase flow thermal-hydraulics allowing easy multi-scale and multi-disciplinary calculations meeting the industrial needs. The NEPTUNE activities include software development, research in physical modeling and numerical methods, the development of advanced instrumentation techniques and performance of new experimental programs. The work focuses on the four different simulation scales: DNS (Direct Numerical Simulation), local CFD (Computational Fluid Dynamics), component (subchannel-type analysis) and system scales. New physical models and numerical methods are being developed for each scale as well as for their coupling. This paper gives an overview of the NEPTUNE activities. It presents the main scientific and technical achievements obtained during Phase 1 (2002-2003) and at the beginning of Phase 2 (2004- 2006). Planned work for the future is also presented. (authors)

  5. NASA's Nuclear Thermal Propulsion Project

    Science.gov (United States)

    Houts, Michael G.; Mitchell, Doyce P.; Kim, Tony; Emrich, William J.; Hickman, Robert R.; Gerrish, Harold P.; Doughty, Glen; Belvin, Anthony; Clement, Steven; Borowski, Stanley K.; hide

    2015-01-01

    The fundamental capability of Nuclear Thermal Propulsion (NTP) is game changing for space exploration. A first generation NTP system 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 a first generation NTP 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 NTP project could also help enable high performance fission power systems and Nuclear Electric Propulsion (NEP).

  6. Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

    International Nuclear Information System (INIS)

    Scari, Maria Elizabeth

    2017-01-01

    Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

  7. Indian advanced nuclear reactors

    International Nuclear Information System (INIS)

    Saha, D.; Sinha, R.K.

    2005-01-01

    For sustainable development of nuclear energy, a number of important issues like safety, waste management, economics etc. are to be addressed. To do this, a number of advanced reactor designs as well as fuel cycle technologies are being pursued worldwide. The advanced reactors being developed in India are the AHWR and the CHTR. Both the reactors use thorium based fuel and have many passive features. This paper describes the Indian advanced reactors and gives a brief account of the international initiatives for the sustainable development of nuclear energy. (author)

  8. Advanced Nuclear Magnetic Resonance

    OpenAIRE

    Alonso, Diego A.

    2014-01-01

    Transparencias en inglés de la asignatura "Resonancia Magnética Nuclear Avanzada" (Advanced Nuclear Magnetic Resonance) (36643) que se imparte en el Máster de Química Médica como asignatura optativa de 3 créditos ECTS. En esta asignatura se completa el estudio iniciado en la asignatura de quinto curso de la licenciatura en Química "Determinación estructural" (7448) y en la del Grado de Química de tercer curso "Determinación estructural de los compuestos orgánicos" (26030) en lo referente a té...

  9. Nuclear imaging: Advances and trends

    International Nuclear Information System (INIS)

    Herk, G. van

    1986-01-01

    In this article, nuclear imaging instruments that are likely to be of interest to the nuclear medicine community of developing countries are emphasized. The advances, trends, developments, and future directions in the field of nuclear imaging are mentioned

  10. TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1984-06-01

    The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

  11. 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

  12. Unique nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Culver, D.W.; Rochow, R.

    1993-06-01

    In January, 1992, a new, advanced nuclear thermal rocket engine (NTRE) concept intended for manned missions to the moon and to Mars was introduced (Culver, 1992). This NTRE promises to be both shorter and lighter in weight than conventionally designed engines, because its forward flowing reactor is located within an expansion-deflection rocket nozzle. The concept has matured during the year, and this paper discusses a nearer term version that resolves four open issues identified in the initial concept: (1) the reactor design and cooling scheme simplification while retaining a high pressure power balance option; (2) elimination need for a new, uncooled nozzle throat material suitable for long life application; (3) a practical provision for reactor power control; and (4) use of near-term, long-life turbopumps

  13. Advanced instrumentation and analysis methods for in-pile thermal and nuclear measurements: from out-of-pile studies to irradiation campaigns

    International Nuclear Information System (INIS)

    Reynard-Carette, C.; Lyoussi, A.

    2015-01-01

    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

  14. 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

  15. 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

  16. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

    International Nuclear Information System (INIS)

    1980-01-01

    Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

  17. Advanced thermally stable jet fuels

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.

    1999-01-31

    The Pennsylvania State University program in advanced thermally stable coal-based jet fuels has five broad objectives: (1) Development of mechanisms of degradation and solids formation; (2) Quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) Characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) Elucidation of the role of additives in retarding the formation of carbonaceous solids; (5) Assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal. Future high-Mach aircraft will place severe thermal demands on jet fuels, requiring the development of novel, hybrid fuel mixtures capable of withstanding temperatures in the range of 400--500 C. In the new aircraft, jet fuel will serve as both an energy source and a heat sink for cooling the airframe, engine, and system components. The ultimate development of such advanced fuels requires a thorough understanding of the thermal decomposition behavior of jet fuels under supercritical conditions. Considering that jet fuels consist of hundreds of compounds, this task must begin with a study of the thermal degradation behavior of select model compounds under supercritical conditions. The research performed by The Pennsylvania State University was focused on five major tasks that reflect the objectives stated above: Task 1: Investigation of the Quantitative Degradation of Fuels; Task 2: Investigation of Incipient Deposition; Task 3: Characterization of Solid Gums, Sediments, and Carbonaceous Deposits; Task 4: Coal-Based Fuel Stabilization Studies; and Task 5: Exploratory Studies on the Direct Conversion of Coal to High Quality Jet Fuels. The major findings of each of these tasks are presented in this executive summary. A description of the sub-tasks performed under each of these tasks and the findings of those studies are provided in the remainder of this volume

  18. Development of demonstration advanced thermal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige

    1982-08-01

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported.

  19. Development of demonstration advanced thermal reactor

    International Nuclear Information System (INIS)

    Nishimura, Seiji; Oguchi, Isao; Touhei, Kazushige.

    1982-01-01

    The design of the advanced thermal demonstration reactor with 600 MWe output was started in 1975. In order to make the compact core, 648 fuel assemblies, each comprising 36 fuel rods, were used, and the mean channel output was increased by 20% as compared with the prototype reactor. The heavy water dumping mechanism for the calandria was abolished. Advanced thermal reactors are suitable to burn plutonium, since the control rod worth does not change, the void reactivity coefficient of coolant shifts to the negative side, and the harmful influence of high order plutonium is small. The void reactivity coefficient is nearly zero, the fluctuation of output in relation to pressure disturbance is small, and the local output change of fuel by the operation of control rods is small, therefore, the operation following load change is relatively easy. The coolant recirculation system is of independent loop construction dividing the core into two, and steam and water are separated in respective steam drums. At present, the rationalizing design is in progress by the leadership of the Power Reactor and Nuclear Fuel Development Corp. The outline of the demonstration reactor, the reactor construction, the nuclear-thermal-hydraulic characteristics and the output control characteristics are reported. (Kako, I.)

  20. Advanced nuclear reactor and nuclear fusion power generation

    International Nuclear Information System (INIS)

    2000-04-01

    This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

  1. Advancing against nuclear terrorism

    International Nuclear Information System (INIS)

    Allison, G.

    2006-01-01

    Meeting a day before the summit, Bush and Putin announced a new Global Initiative to Combat Nuclear Terrorism; a plan for multiple, multilateral guaranteed suppliers of nuclear fuel to States that forgo building their own enrichment plants; and a Civil Nuclear Agreement that will lift restrictions on cooperation between the two countries in developing peaceful nuclear power. Each of these initiatives provides a framework for dozens of specific actions that can measurably reduce the risk of terrorists acquiring a nuclear weapon. The significance of the Global Initiative against Nuclear Terrorism lies not only in its substance but in Russia's visible joint ownership of the Initiative. After years in which Washington lectured Moscow about this threat, Putin's joint leadership in securing nuclear material worldwide should give added impetus to this undertaking inside Russia as well. Globally, this initiative calls for work plans in five arenas: prevention, detection, disruption, mitigation of consequences after an attack, and strengthening domestic laws and export controls against future A.Q. Khans. The guaranteed nuclear fuel supply tightens the noose around Iran as it seeks to exploit a loophole in the global Nuclear Non-Proliferation Treaty. By guaranteeing States that six separate international suppliers will provide backup guarantees against interruption of supply for any reason other that breech of commitments under the NPT, this proposal eliminates Iran's excuse for Natanz-the enrichment plant it is rushing to finish today. This system for supply will be subject to the supervision by the IAEA, which will also have nuclear fuel reserves that allow it to be a supplier of last resort. The Civil Nuclear Agreement will allow joint research on next-generation, proliferation-proof reactors, including technologies where Russian science is the best in the world. It will permit sale to Russia of US technologies that can improve the safety and efficiency of Russian nuclear

  2. 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...

  3. Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

    1994-07-01

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  4. Computational Design of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

    2014-01-01

    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.

  5. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    Berglund, R.C.

    1993-01-01

    The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

  6. A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

    Energy Technology Data Exchange (ETDEWEB)

    D’Auria, F; Rohatgi, Upendra S.

    2017-01-12

    The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

  7. Nuclear thermal propulsion workshop overview

    International Nuclear Information System (INIS)

    Clark, J.S.

    1991-01-01

    NASA is planning an Exploration Technology Program as part of the Space Exploration Initiative to return U.S. astronauts to the moon, conduct intensive robotic exploration of the moon and Mars, and to conduct a piloted mission to Mars by 2019. Nuclear Propulsion is one of the key technology thrust for the human mission to Mars. The workshop addresses NTP (Nuclear Thermal Rocket) technologies with purpose to: assess the state-of-the-art of nuclear propulsion concepts; assess the potential benefits of the concepts for the mission to Mars; identify critical, enabling technologies; lay-out (first order) technology development plans including facility requirements; and estimate the cost of developing these technologies to flight-ready status. The output from the workshop will serve as a data base for nuclear propulsion project planning

  8. Advances in nuclear desalination

    International Nuclear Information System (INIS)

    Misra, B.M.

    2003-01-01

    The Nuclear Desalination Demonstration Project (NDDP) at Kalpakkam aims to demonstrate the safe and economic production of good quality water by desalination of seawater comprising 4,500 m 3 /d Multi-Stage Flash (MSF) and 1,800 m 3 /d Reverse Osmosis (RO) plant. The design of the hybrid MSF-RO plant to be set up at an existing nuclear power station is presented. The MSF plant based on long tube design requires less energy. The effect on performance of the MSF plant due to higher seawater intake temperature is marginal. The preheat RO system part of the hybrid plant uses reject cooling seawater from the MSF plant. This allows lower pressure operation, resulting in energy saving. The two qualities of water produced are usable for the power station as well as for drinking purposes with appropriate blending. The post treatment is also simplified due to blending of the products from MSF and RO plants. The hybrid plant has a number of advantages: part of high purity desalted water produced from the MSF plant will be used for the makeup demineralised water requirement (after necessary polishing) for the power station; blending of the product water from RO and MSF plants would provide requisite quality drinking water; and the RO plant will continue to be operated to provide water for drinking purposes during the shut down of the power station. Commissioning of the RO section is expected in 2002 and that of the MSF section in 2003. Useful design data are expected from the plant on the coupling of small and medium size reactors (SMR) based on PHWR. This will enable us to design a large size commercial plant up to 50,000 m 3 /d capacity. India will share the O and M experience of NDDP to member states of the International Atomic Energy Agency (IAEA) when the plant is commissioned. The development work for producing good quality water for power station from high salinity water utilizing low grade waste heat is presented. About 40 and 100 MWth low temperature waste heat is

  9. 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...

  10. Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D; Estudo termofluidodinâmico de reatores nucleares avançados de alta temperatura utilizando o RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Scari, Maria Elizabeth

    2017-07-01

    Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF{sub 2}, the LiF-BeF{sub 2}, also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO

  11. Advanced teleoperation in nuclear applications

    International Nuclear Information System (INIS)

    Hamel, W.R.; Feldman, M.J.; Martin, H.L.

    1984-01-01

    A new generation of integrated remote maintenance systems is being developed to meet the needs of future nuclear fuel reprocessing at the Oak Ridge National Laboratory. Development activities cover all aspects of an advanced teleoperated maintenance system with particular emphasis on a new force-reflecting servomanipulator concept. The new manipulator, called the advanced servomanipulator, is microprocessor controlled and is designed to achieve force-reflection performance near that of mechanical master/slave manipulators. The advanced servomanipulator uses a gear-drive transmission which permits modularization for remote maintainability (by other advanced servomanipulators) and increases reliability. Human factors analysis has been used to develop an improved man/machine interface concept based upon colorgraphic displays and menu-driven tough screens. Initial test and evaluation of two advanced servomanipulator slave arms and several other development components have begun. 9 references, 5 figures

  12. 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.

  13. Advanced nuclear precleaner

    International Nuclear Information System (INIS)

    Wright, S.R.

    1997-01-01

    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

  14. Recent advances in nuclear cardiology

    International Nuclear Information System (INIS)

    Lee, Won Woo

    2016-01-01

    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

  15. Nuclear Thermal Propulsion Development Risks

    Science.gov (United States)

    Kim, Tony

    2015-01-01

    There are clear advantages of development of a Nuclear Thermal Propulsion (NTP) for a crewed mission to Mars. NTP for in-space propulsion enables more ambitious space missions by providing high thrust at high specific impulse ((is) approximately 900 sec) that is 2 times the best theoretical performance possible for chemical rockets. Missions can be optimized for maximum payload capability to take more payload with reduced total mass to orbit; saving cost on reduction of the number of launch vehicles needed. Or missions can be optimized to minimize trip time significantly to reduce the deep space radiation exposure to the crew. NTR propulsion technology is a game changer for space exploration to Mars and beyond. However, 'NUCLEAR' is a word that is feared and vilified by some groups and the hostility towards development of any nuclear systems can meet great opposition by the public as well as from national leaders and people in authority. The public often associates the 'nuclear' word with weapons of mass destruction. The development NTP is at risk due to unwarranted public fears and clear honest communication of nuclear safety will be critical to the success of the development of the NTP technology. Reducing cost to NTP development is critical to its acceptance and funding. In the past, highly inflated cost estimates of a full-scale development nuclear engine due to Category I nuclear security requirements and costly regulatory requirements have put the NTP technology as a low priority. Innovative approaches utilizing low enriched uranium (LEU). Even though NTP can be a small source of radiation to the crew, NTP can facilitate significant reduction of crew exposure to solar and cosmic radiation by reducing trip times by 3-4 months. Current Human Mars Mission (HMM) trajectories with conventional propulsion systems and fuel-efficient transfer orbits exceed astronaut radiation exposure limits. Utilizing extra propellant from one additional SLS launch and available

  16. Advanced nuclear systems. Review study

    International Nuclear Information System (INIS)

    Liebert, Wolfgang; Glaser, Alexander; Pistner, Christoph; Baehr, Roland; Hahn, Lothar

    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

  17. Advanced technology for nuclear powerplants

    International Nuclear Information System (INIS)

    Rohm, H.H.

    1987-01-01

    Advanced technology offers significant potential benefit to the nuclear industry. Improvements can be anticipated in plant performance, reliability, and overall plant safety as well as reduced life cycle costs. Utilizing artificial intelligence and expert systems, robotics, advanced instruments and controls, and modularization technologies can enhance plant operations and provide new insights and perspectives to plant risk and thus focus resources to areas of importance. Plant reliability, operability, availability, accident interdiction and limitation, and plant recovery are expected to improve. However, utilizing these technologies is not an automatic process. In addition to the actual costs associated with developing and implementing the technologies, operator training and acceptance represents a potential significant problem. Traditional plant operators have little or no experience with computer technology. There has already been some difficulty getting nuclear plant operators to accept and use the new technologies that have been implemented to accept and use the new technologies that have been implemented thus far

  18. Advanced nuclear energy analysis technology

    International Nuclear Information System (INIS)

    Gauntt, Randall O.; Murata, Kenneth K.; Romero, Vicente Josce; Young, Michael Francis; Rochau, Gary Eugene

    2004-01-01

    A two-year effort focused on applying ASCI technology developed for the analysis of weapons systems to the state-of-the-art accident analysis of a nuclear reactor system was proposed. The Sandia SIERRA parallel computing platform for ASCI codes includes high-fidelity thermal, fluids, and structural codes whose coupling through SIERRA can be specifically tailored to the particular problem at hand to analyze complex multiphysics problems. Presently, however, the suite lacks several physics modules unique to the analysis of nuclear reactors. The NRC MELCOR code, not presently part of SIERRA, was developed to analyze severe accidents in present-technology reactor systems. We attempted to: (1) evaluate the SIERRA code suite for its current applicability to the analysis of next generation nuclear reactors, and the feasibility of implementing MELCOR models into the SIERRA suite, (2) examine the possibility of augmenting ASCI codes or alternatives by coupling to the MELCOR code, or portions thereof, to address physics particular to nuclear reactor issues, especially those facing next generation reactor designs, and (3) apply the coupled code set to a demonstration problem involving a nuclear reactor system. We were successful in completing the first two in sufficient detail to determine that an extensive demonstration problem was not feasible at this time. In the future, completion of this research would demonstrate the feasibility of performing high fidelity and rapid analyses of safety and design issues needed to support the development of next generation power reactor systems

  19. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and 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 enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via 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. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

  20. Advanced thermal management technologies for defense electronics

    Science.gov (United States)

    Bloschock, Kristen P.; Bar-Cohen, Avram

    2012-05-01

    Thermal management technology plays a key role in the continuing miniaturization, performance improvements, and higher reliability of electronic systems. For the past decade, and particularly, the past 4 years, the Defense Advanced Research Projects Agency (DARPA) has aggressively pursued the application of micro- and nano-technology to reduce or remove thermal constraints on the performance of defense electronic systems. The DARPA Thermal Management Technologies (TMT) portfolio is comprised of five technical thrust areas: Thermal Ground Plane (TGP), Microtechnologies for Air-Cooled Exchangers (MACE), NanoThermal Interfaces (NTI), Active Cooling Modules (ACM), and Near Junction Thermal Transport (NJTT). An overview of the TMT program will be presented with emphasis on the goals and status of these efforts relative to the current State-of-the-Art. The presentation will close with future challenges and opportunities in the thermal management of defense electronics.

  1. Philosophy for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Buden, D.; Madsen, W.; Redd, L.

    1993-01-01

    The philosophy used for development of nuclear thermal propulsion will determine the cost, schedule and risk associated with the activities. As important is the impression of the decision makers. If the development cost is higher than the product value, it is doubtful that funding will ever be available. On the other hand, if the development supports the economic welfare of the country with a high rate of return, the probability of funding greatly increases. The philosophy is divided into: realism, design, operations and qualification. ''Realism'' addresses such items as political acceptability, potential customers, robustness-flexibility, public acceptance, decisions as needed, concurrent engineering, and the possible role of the CIS. ''Design'' addresses ''minimum requirement,'' built in safety and reliability redundancy, emphasize on eliminating risk at lowest levels, and the possible inclusion of electric generation. ''Operations'' addresses sately, environment, operations, design margins and degradation modes. ''Qualification'' addresses testing needs and test facilities

  2. Advanced nuclear plant control complex

    International Nuclear Information System (INIS)

    Scarola, K.; Jamison, S.; Manazir, R.M.; Rescorl, R.L.; Harmon, D.L.

    1991-01-01

    An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system which is nuclear qualified for rapid response to changes in plant parameters and a component control system which together provide a discrete monitoring and control capability at a panel in the control room. A separate data processing system, which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs and a large, overhead integrated process status overview board. The discrete indicator and alarm system and the data processing system receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accidental conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof. (author)

  3. Lunar mission design using nuclear thermal rockets

    International Nuclear Information System (INIS)

    Stancati, M.L.; Collins, J.T.; Borowski, S.K.

    1991-01-01

    The NERVA-class Nuclear Thermal Rocket (NTR), with performance nearly double that of advanced chemical engines, has long been considered an enabling technology for human missions to Mars. NTR engines address the demanding trip time and payload delivery needs of both cargo-only and piloted flights. But NTR can also reduce the Earth launch requirements for manned lunar missions. First use of NTR for the Moon would be less demanding and would provide a test-bed for early operations experience with this powerful technology. Study of application and design options indicates that NTR propulsion can be integrated with the Space Exploration Initiative scenarios to deliver performance gains while managing controlled, long-term disposal of spent reactors to highly stable orbits

  4. Space Nuclear Thermal Propulsion (SNTP) Air Force facility

    Science.gov (United States)

    Beck, David F.

    The Space Nuclear Thermal Propulsion (SNTP) Program is an initiative within the US Air Force to acquire and validate advanced technologies that could be used to sustain superior capabilities in the area or space nuclear propulsion. The SNTP Program has a specific objective of demonstrating the feasibility of the particle bed reactor (PBR) concept. The term PIPET refers to a project within the SNTP Program responsible for the design, development, construction, and operation of a test reactor facility, including all support systems, that is intended to resolve program technology issues and test goals. A nuclear test facility has been designed that meets SNTP Facility requirements. The design approach taken to meet SNTP requirements has resulted in a nuclear test facility that should encompass a wide range of nuclear thermal propulsion (NTP) test requirements that may be generated within other programs. The SNTP PIPET project is actively working with DOE and NASA to assess this possibility.

  5. Advanced nuclear systems in comparison

    International Nuclear Information System (INIS)

    Brogli, R.; Foskolos, K.; Goetzmann, C.; Kroeger, W.; Stanculescu, A.; Wydler, P.

    1996-09-01

    This study aims at a comparison of future reactor concepts, paying particular attention to aspects of safety, of the fuel cycle, the economics, the experience-base and the state of development. Representative examples of typical development lines, that could possibly be 'of interest' within a time horizon of 50 years were selected for comparison. This can be divided into three phases: - Phase I includes the next 10 years and will be characterised mainly by evolutionary developments of light water reactors (LWR) of large size; representative: EPR, - Phase II: i.e. the time between 2005 and 2020 approximately, encompasses the forecasted doubling of today's world-wide installed nuclear capacity; along with evolutionary reactors, innovative systems like AP600, PIUS, MHTGR, EFR will emerge, - Phase III covers the time between 2020 and 2050 and is characterised by the issue of sufficient fissile material resources; novel fast reactor systems including hybrid systems can, thus, become available; representatives: IFR, EA, ITER (the latter being). The evaluated concepts foresee partly different fuel cycles. Fission reactors can be operated in principle on the basis of either a Uranium-Plutonium-cycle or a Thorium-Uranium-cycle, while combinations of these cycles among them or with other reactor concepts than proposed are possible. With today's nuclear park (comprising mainly LWRs), the world-wide plutonium excess increases annually by about 100 t. Besides strategies based on reprocessing like: - recycling in thermal and fast reactors with mixed oxide fuels, - plutonium 'burning' in reactors with novel fuels without uranium or in 'hybrid' systems, allowing a reduction of this excess, direct disposal of spent fuel elements including their plutonium content ('one-through') is being considered. (author) figs., tabs., 32 refs

  6. Development of a 1200 fine group nuclear data library for advanced nuclear systems

    Institute of Scientific and Technical Information of China (English)

    Jun Zou; Lei-Ming Shang; Fang Wang; Li-Juan Hao

    2017-01-01

    Accurate and reliable nuclear data libraries are essential for calculation and design of advanced nuclear systems.A 1200 fine group nuclear data library Hybrid Evaluated Nuclear Data Library/Fine Group (HENDL/FG) with neutrons of up to 150 MeV has been developed to improve the accuracy of neutronics calculations and analysis.Corrections of Doppler,resonance self-shielding,and thermal upscatter effects were done for HENDL/FG.Shielding and critical safety benchmarks were performed to test the accuracy and reliability of the library.The discrepancy between calculated and measured nuclear parameters fell into a reasonable range.

  7. Current Comparison of Advanced Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    Steven Piet; Trond Bjornard; Brent Dixon; Robert Hill; Gretchen Matthern; David Shropshire

    2007-01-01

    This paper compares potential nuclear fuel cycle strategies--once-through, recycling in thermal reactors, sustained recycle with a mix of thermal and fast reactors, and sustained recycle with fast reactors. Initiation of recycle starts the draw-down of weapons-usable material and starts accruing improvements for geologic repositories and energy sustainability. It reduces the motivation to search for potential second geologic repository sites. Recycle in thermal-spectrum nuclear reactors achieves several recycling objectives; fast nuclear reactors achieve all of them

  8. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  9. Evaluation and development of advanced nuclear materials: IAEA activities

    International Nuclear Information System (INIS)

    Inozemtsev, V.; Basak, U.; Killeen, J.; Dyck, G.; Zeman, A.; )

    2011-01-01

    Economical, environmental and non-proliferation issues associated with sustainable development of nuclear power bring about a need for optimization of fuel cycles and implementation of advanced nuclear systems. While a number of physical and design concepts are available for innovative reactors, the absence of reliable materials able to sustain new challenging irradiation conditions represents the real bottle-neck for practical implementation of these promising ideas. Materials performance and integrity are key issues for the safety and competitiveness of future nuclear installations being developed for sustainable nuclear energy production incorporating fuel recycling and waste transmutation systems. These systems will feature high thermal operational efficiency, improved utilization of resources (both fissile and fertile materials) and reduced production of nuclear waste. They will require development, qualification and deployment of new and advanced fuel and structural materials with improved mechanical and chemical properties combined with high radiation and corrosion resistance. The extensive, diverse, and expensive efforts toward the development of these materials can be more effectively organized within international collaborative programmes with wide participation of research, design and engineering communities. IAEA carries out a number of international projects supporting interested Member States with the use of available IAEA program implementation tools (Coordinated Research Projects, Technical Meetings, Expert Reviews, etc). The presentation summarizes the activities targeting material developments for advanced nuclear systems, with particular emphasis on fast reactors, which are the focal topics of IAEA Coordinated Research Projects 'Accelerator Simulation and Theoretical Modelling of Radiation Effects' (on-going), 'Benchmarking of Structural Materials Pre-Selected for Advanced Nuclear Reactors', 'Examination of advanced fast reactor fuel and core

  10. Proceedings of the second international conference on advances in nuclear materials: abstract booklet and souvenir

    International Nuclear Information System (INIS)

    2011-01-01

    Nuclear materials form special class of materials which either act as fuel for the nuclear reactors or form the structure of the reactors and the allied systems. The topics covered in this conference are: materials challenges for thermal and fast reactors, technological advances in nuclear fuels and components, materials for future reactors, fuel cycles and materials challenges, materials degradation and life management, advanced materials development, modelling and simulation, advanced materials- II, advanced materials for future reactors, development of advanced fuel and structural materials, zirconium alloy developments, irradiation effects and PIE, advanced nuclear fuels, corrosion and materials characterization. Papers relevant to INIS are indexed separately

  11. Thermal modelling of Advanced LIGO test masses

    International Nuclear Information System (INIS)

    Wang, H; Dovale Álvarez, M; Mow-Lowry, C M; Freise, A; Blair, C; Brooks, A; Kasprzack, M F; Ramette, J; Meyers, P M; Kaufer, S; O’Reilly, B

    2017-01-01

    High-reflectivity fused silica mirrors are at the epicentre of today’s advanced gravitational wave detectors. In these detectors, the mirrors interact with high power laser beams. As a result of finite absorption in the high reflectivity coatings the mirrors suffer from a variety of thermal effects that impact on the detectors’ performance. We propose a model of the Advanced LIGO mirrors that introduces an empirical term to account for the radiative heat transfer between the mirror and its surroundings. The mechanical mode frequency is used as a probe for the overall temperature of the mirror. The thermal transient after power build-up in the optical cavities is used to refine and test the model. The model provides a coating absorption estimate of 1.5–2.0 ppm and estimates that 0.3 to 1.3 ppm of the circulating light is scattered onto the ring heater. (paper)

  12. Revision of construction plan for advanced thermal demonstration reactor

    International Nuclear Information System (INIS)

    1996-01-01

    The Federation of Electric Power Companies demanded the revision of the construction plan for the advanced thermal demonstration reactor, which is included in the 'Long term plan on the research, development and utilization of atomic energy' decided by the Atomic Energy Commission in 1994, for economical reason. The Atomic Energy Commission carried out the deliberation on this demand. It was found that the cost of construction increases to 580 billion yen, and the cost of electric power generation increases three times as high as that of LWRs. The role as the reactor that utilizes MOX fuel can be substituted by LWRs. The relation of trust with the local town must be considered. In view of these circumstances, it is judged that the stoppage of the construction plan is appropriate. It is necessary to investigate the substitute plan for the stoppage, and the viewpoints of investigating the substitute plan, the examination of the advanced BWR with all MOX fuel core and the method of advancing its construction are considered. On the research and development related to advanced thermal reactors, the research and development contributing to the advance of nuclear fuel recycling are advanced, and the prototype reactor 'Fugen' is utilized. (K.I.)

  13. Thermal performance and efficiency of supercritical nuclear reactors

    International Nuclear Information System (INIS)

    Romney Duffey; Tracy Zhou; Hussam Khartabil

    2009-01-01

    The paper reviews the major advances and innovative aspects of the thermal performance of recent concepts for super-critical water-cooled nuclear reactors (SCWR). The concepts are based on the extensive experience in the thermal power industry with super and ultra-supercritical boilers and turbines. The challenges and goals of increased efficiency, reduced cost, enhanced safety and co-generation have been pursued over the last ten years, and have resulted both in viable concepts and a vibrant defined R and D effort. The supercritical concept has wide acceptance among industry, as it reflects standard engineering practices and current thermal plant technology that is being already deployed. The SCWR concept represents a continuous development of water-cooled reactor technology, which utilizes the best and latest advances made in the thermal power industry. (author)

  14. Thermal hydraulics in undergraduate nuclear engineering education

    International Nuclear Information System (INIS)

    Theofanous, T.G.

    1986-01-01

    The intense safety-related research efforts of the seventies in reactor thermal hydraulics have brought about the recognition of the subject as one of the cornerstones of nuclear engineering. Many nuclear engineering departments responded by building up research programs in this area, and mostly as a consequence, educational programs, too. Whether thermal hydraulics has fully permeated the conscience of nuclear engineering, however, remains yet to be seen. The lean years that lie immediately ahead will provide the test. The purpose of this presentation is to discuss the author's own educational activity in undergraduate nuclear engineering education over the past 10 yr or so. All this activity took place at Purdue's School of Nuclear Engineering. He was well satisfied with the results and expects to implement something similar at the University of California in Santa Barbara in the near future

  15. Thermal-hydraulic analysis of nuclear reactors

    CERN Document Server

    Zohuri, Bahman

    2015-01-01

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

  16. Thermal-hydraulic-geochemical coupled processes around disposed high level nuclear waste in deep granite hosted geological repositories: frontier areas of advanced groundwater research in India

    International Nuclear Information System (INIS)

    Bajpai, R.K.

    2012-01-01

    Indian policy for permanent disposal of high level nuclear wastes with radionuclide having very long half lives include their immobilization in a stable matrix i.e. glasses of suitable composition, its storage in high integrity steel canisters and subsequent disposal in suitable host rock like granites at a depth of 400-500m in stable geological set up. The site for such disposal facilities are selected after vigorous assessment of their stability implying an exhaustive site selection methodology based on a large number of criteria and attributes. In India, an area of about 70000 square kilometers occupied by granites has been subjected to such evaluation for generating comprehensive database on host rock parameters. The sites selected after such intensive analysis are expected to remain immune to processes like seismicity, volcanism, faulting, uplift, erosion, flooding etc. even in distant future spanning over tens of thousands of years. Nevertheless, groundwater has emerged as the only credible pathway through which disposed waste can eventually find its way to accessible biosphere. Hence groundwater research constitutes one of the most important aspects in demonstration of safety of such disposal. The disposed waste due to continuous emission of decay heat creates high temperature field around them with resultant increase in groundwater temperature in the vicinity. Hot groundwater on reacting with steel canisters, backfill clays and cement used around the disposed canister, produces geochemical environment characterized by altered Ph, Eh and groundwater compositions. Acceleration in geochemical interaction among waste-groundwater-clay-cement-granite often results in dissolution or precipitation reactions along the groundwater flow paths i.e. fractures with resultant increase or decrease in their permeability. Thus thermal, hydraulic and geochemical processes work interdependently around the disposed waste. These coupled processes also control the release and

  17. Advanced materials for thermal protection system

    Science.gov (United States)

    Heng, Sangvavann; Sherman, Andrew J.

    1996-03-01

    Reticulated open-cell ceramic foams (both vitreous carbon and silicon carbide) and ceramic composites (SiC-based, both monolithic and fiber-reinforced) were evaluated as candidate materials for use in a heat shield sandwich panel design as an advanced thermal protection system (TPS) for unmanned single-use hypersonic reentry vehicles. These materials were fabricated by chemical vapor deposition/infiltration (CVD/CVI) and evaluated extensively for their mechanical, thermal, and erosion/ablation performance. In the TPS, the ceramic foams were used as a structural core providing thermal insulation and mechanical load distribution, while the ceramic composites were used as facesheets providing resistance to aerodynamic, shear, and erosive forces. Tensile, compressive, and shear strength, elastic and shear modulus, fracture toughness, Poisson's ratio, and thermal conductivity were measured for the ceramic foams, while arcjet testing was conducted on the ceramic composites at heat flux levels up to 5.90 MW/m2 (520 Btu/ft2ṡsec). Two prototype test articles were fabricated and subjected to arcjet testing at heat flux levels of 1.70-3.40 MW/m2 (150-300 Btu/ft2ṡsec) under simulated reentry trajectories.

  18. 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

  19. ATLAS program for advanced thermal-hydraulic safety research

    International Nuclear Information System (INIS)

    Song, Chul-Hwa; Choi, Ki-Yong; Kang, Kyoung-Ho

    2015-01-01

    Highlights: • Major achievements of the ATLAS program are highlighted in conjunction with both developing advanced light water reactor technologies and enhancing the nuclear safety. • The ATLAS data was shown to be useful for the development and licensing of new reactors and safety analysis codes, and also for nuclear safety enhancement through domestic and international cooperative programs. • A future plan for the ATLAS testing is introduced, covering recently emerging safety issues and some generic thermal-hydraulic concerns. - Abstract: This paper highlights the major achievements of the ATLAS program, which is an integral effect test program for both developing advanced light water reactor technologies and contributing to enhancing nuclear safety. The ATLAS program is closely related with the development of the APR1400 and APR"+ reactors, and the SPACE code, which is a best-estimate system-scale code for a safety analysis of nuclear reactors. The multiple roles of ATLAS testing are emphasized in very close conjunction with the development, licensing, and commercial deployment of these reactors and their safety analysis codes. The role of ATLAS for nuclear safety enhancement is also introduced by taking some examples of its contributions to voluntarily lead to multi-body cooperative programs such as domestic and international standard problems. Finally, a future plan for the utilization of ATLAS testing is introduced, which aims at tackling recently emerging safety issues such as a prolonged station blackout accident and medium-size break LOCA, and some generic thermal-hydraulic concerns as to how to figure out multi-dimensional phenomena and the scaling issue.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. A development approach for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Buden, D.

    1992-01-01

    The cost and time to develop nuclear thermal propulsion systems are very approach dependent. The objectives addressed are the development of an ''acceptable'' nuclear thermal propulsion system that can be used as part of the transportation system for people to explore Mars and the enhancement performance of other missions, within highly constrained budgets and schedules. To accomplish this, it was necessary to identify the cost drivers considering mission parameters, safety of the crew, mission success, facility availability and time and cost to construct new facilities, qualification criteria, status of technologies, management structure, and use of such system engineering techniques as concurrent engineering

  2. Nuclear thermal rockets using indigenous Martian propellants

    International Nuclear Information System (INIS)

    Zubrin, R.M.

    1989-01-01

    This paper considers a novel concept for a Martian descent and ascent vehicle, called NIMF (for nuclear rocket using indigenous Martian fuel), the propulsion for which will be provided by a nuclear thermal reactor which will heat an indigenous Martian propellant gas to form a high-thrust rocket exhaust. The performance of each of the candidate Martian propellants, which include CO2, H2O, CH4, N2, CO, and Ar, is assessed, and the methods of propellant acquisition are examined. Attention is also given to the issues of chemical compatibility between candidate propellants and reactor fuel and cladding materials, and the potential of winged Mars supersonic aircraft driven by this type of engine. It is shown that, by utilizing the nuclear landing craft in combination with a hydrogen-fueled nuclear thermal interplanetary vehicle and a heavy lift booster, it is possible to achieve a manned Mars mission in one launch. 6 refs

  3. 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

  4. Robot vision for nuclear advanced robot

    International Nuclear Information System (INIS)

    Nakayama, Ryoichi; Okano, Hideharu; Kuno, Yoshinori; Miyazawa, Tatsuo; Shimada, Hideo; Okada, Satoshi; Kawamura, Astuo

    1991-01-01

    This paper describes Robot Vision and Operation System for Nuclear Advanced Robot. This Robot Vision consists of robot position detection, obstacle detection and object recognition. With these vision techniques, a mobile robot can make a path and move autonomously along the planned path. The authors implemented the above robot vision system on the 'Advanced Robot for Nuclear Power Plant' and tested in an environment mocked up as nuclear power plant facilities. Since the operation system for this robot consists of operator's console and a large stereo monitor, this system can be easily operated by one person. Experimental tests were made using the Advanced Robot (nuclear robot). Results indicate that the proposed operation system is very useful, and can be operate by only person. (author)

  5. Macroscopic dynamics of thermal nuclear excitations

    International Nuclear Information System (INIS)

    Bastrukov, S.I.; Deak, F.; Kiss, A.; Seres, Z.

    1989-11-01

    The concept of kinetic temperature as a local dynamical variable of thermal nuclear collective motion is formulated using long-mean-free-path approach based on the Landau-Vlasov kinetic equation. In the Fermi drop model the thermal fluid dynamics of the spherical nucleus is analyzed. It is shown that in a compressible Fermi liquid the temperature pulses propagate in the form of spherical wave in phase with the acoustic wave. The thermal and compressional excitations are caused by the isotropic harmonic oscillations of the Fermi sphere in momentum space. (author) 25 refs.; 2 figs

  6. Evaluation of Erosion Resistance of Advanced Turbine Thermal Barrier Coatings

    Science.gov (United States)

    Zhu, Dongming; Kuczmarski, Maria A.; Miller, Robert A.; Cuy, Michael D.

    2007-01-01

    The erosion resistant turbine thermal barrier coating system is critical to aircraft engine performance and durability. By demonstrating advanced turbine material testing capabilities, we will be able to facilitate the critical turbine coating and subcomponent development and help establish advanced erosion-resistant turbine airfoil thermal barrier coatings design tools. The objective of this work is to determine erosion resistance of advanced thermal barrier coating systems under simulated engine erosion and/or thermal gradient environments, validating advanced turbine airfoil thermal barrier coating systems based on nano-tetragonal phase toughening design approaches.

  7. Advances in thermal-hydraulic studies of a transmutation advanced device for sustainable energy applications

    International Nuclear Information System (INIS)

    Fajardo, Laura Garcia; Castells, Facundo Alberto Escriva; Lira, Carlos Brayner de Olivera

    2013-01-01

    The Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) is a pebble-bed Accelerator Driven System (ADS) with a graphite-gas configuration, designed for nuclear waste trans- mutation and for obtaining heat at very high temperatures to produce hydrogen. In previous work, the TADSEA's nuclear core was considered as a porous medium performed with a CFD code and thermal-hydraulic studies of the nuclear core were presented. In this paper, the heat transfer from the fuel to the coolant was analyzed for three core states during normal operation. The heat transfer inside the spherical fuel elements was also studied. Three critical fuel elements groups were defined regarding their position inside the core. Results were compared with a realistic CFD model of the critical fuel elements groups. During the steady state, no critical elements reached the limit temperature of this type of fuel. (author)

  8. 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.

  9. Research advancements and applications of carboranes in nuclear medicinal chemistry

    International Nuclear Information System (INIS)

    Chen Wen; Wei Hongyuan; Luo Shunzhong

    2011-01-01

    Because of their uniquely high thermal and chemical stabilities, carboranes have become a subject of study with high interest in the chemistry of supra molecules, catalysts and radiopharmaceuticals. In recent years, the role of carboranes in nuclear medicinal chemistry has been diversified, from the traditional use in boron neutron capture therapy (BNCT), to the clinical applications in molecular radio imaging and therapy. This paper provides an overview of the synthesis and characterization of carboranes and their applications in nuclear medicinal chemistry, with highlights of recent key advancements in the re- search areas of BNCT and radio imaging. (authors)

  10. An historical collection of papers on nuclear thermal propulsion

    Science.gov (United States)

    The present volume of historical papers on nuclear thermal propulsion (NTP) encompasses NTP technology development regarding solid-core NTP technology, advanced concepts from the early years of NTP research, and recent activities in the field. Specific issues addressed include NERVA rocket-engine technology, the development of nuclear rocket propulsion at Los Alamos, fuel-element development, reactor testing for the Rover program, and an overview of NTP concepts and research emphasizing two decades of NASA research. Also addressed are the development of the 'nuclear light bulb' closed-cycle gas core and a demonstration of a fissioning UF6 gas in an argon vortex. The recent developments reviewed include the application of NTP to NASA's Lunar Space Transportation System, the use of NTP for the Space Exploration Initiative, and the development of nuclear rocket engines in the former Soviet Union.

  11. National nuclear power planning of China and advanced reactor

    International Nuclear Information System (INIS)

    Qian Jihui

    1990-01-01

    The necessity of investigation on the trends of advanced reactor technology all over the world is elabrated while China is going to set up its long-term national nuclear power programme. In author's opinion, thermal reactor power plants will have a quite long period development in the next century and a new trend of second generation NPPs might emerge in the beginning of next century. These new generation advanced reactors are characterized with new design concepts based on the inherent or passive safety features. Among them, most promising ones are those of AP-600 and MHTGR. Chinese experts are paying special attention to and closely following these two directions

  12. Coordinating Space Nuclear Research Advancement and Education

    International Nuclear Information System (INIS)

    Bess, John D.; Webb, Jonathon A.; Gross, Brian J.; Craft, Aaron E.

    2009-01-01

    The advancement of space exploration using nuclear science and technology has been a goal sought by many individuals over the years. The quest to enable space nuclear applications has experienced many challenges such as funding restrictions; lack of political, corporate, or public support; and limitations in educational opportunities. The Center for Space Nuclear Research (CSNR) was established at the Idaho National Laboratory (INL) with the mission to address the numerous challenges and opportunities relevant to the promotion of space nuclear research and education.1 The CSNR is operated by the Universities Space Research Association and its activities are overseen by a Science Council comprised of various representatives from academic and professional entities with space nuclear experience. Program participants in the CSNR include academic researchers and students, government representatives, and representatives from industrial and corporate entities. Space nuclear educational opportunities have traditionally been limited to various sponsored research projects through government agencies or industrial partners, and dedicated research centers. Centralized research opportunities are vital to the growth and development of space nuclear advancement. Coordinated and focused research plays a key role in developing the future leaders in the space nuclear field. The CSNR strives to synchronize research efforts and provide means to train and educate students with skills to help them excel as leaders.

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

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    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 −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

  14. 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

  15. Advanced Nuclear Fuels Corporation: one year later

    International Nuclear Information System (INIS)

    Bjoernard, T.A.; Sofer, G.A.

    1988-01-01

    About one year ago, after 18 years of business as a wholly owned affiliate of Exxon Corporation, Exxon Nuclear Company was acquired by Siemens/KWU and its name was changed to Advanced Nuclear Fuels Corporation (ANF). This profile describes the status of ANF one year later, principally from the European perspective but with some mention of ANF's worldwide operations to provide a balanced picture. After one year of operation as an affiliate of Siemens/KWU, ANF's role remains as an independent international supplier of nuclear fuel and services to utilities in Europe, the USA and the Far East, but with substantially augmented capabilities resulting from the new affiliation

  16. Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    International Nuclear Information System (INIS)

    Emrich, William J. Jr.

    2008-01-01

    To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts

  17. Nuclear power and other thermal power

    International Nuclear Information System (INIS)

    Bakke, J.

    1978-01-01

    Some philosophical aspects of mortality statistics are first briefly mentioued, then the environmental problems of, first, nuclear power plants, then fossil fuelled power plants are summarised. The effects of releases of carbon dioxide, sulphur dioxide and nitrogen oxides are briefly discussed. The possible health effects of radiation from nuclear power plants and those of gaseous and particulate effluents from fossil fuel plants are also discussed. It is pointed out that in choosing between alternative evils the worst course is to make no choice at all, that is, failure to install thermal power plants will lead to isolated domestic burning of fossil fuels which is clearly the worst situation regarding pollution. (JIW)

  18. 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

  19. Bimodal Nuclear Thermal Rocket Analysis Developments

    Science.gov (United States)

    Belair, Michael; Lavelle, Thomas; Saimento, Charles; Juhasz, Albert; Stewart, Mark

    2014-01-01

    Nuclear thermal propulsion has long been considered an enabling technology for human missions to Mars and beyond. One concept of operations for these missions utilizes the nuclear reactor to generate electrical power during coast phases, known as bimodal operation. This presentation focuses on the systems modeling and analysis efforts for a NERVA derived concept. The NERVA bimodal operation derives the thermal energy from the core tie tube elements. Recent analysis has shown potential temperature distributions in the tie tube elements that may limit the thermodynamic efficiency of the closed Brayton cycle used to generate electricity with the current design. The results of this analysis are discussed as well as the potential implications to a bimodal NERVA type reactor.

  20. Materials technologies for advanced nuclear energy concepts

    International Nuclear Information System (INIS)

    DiStefano, J.; Harms, B.

    1983-01-01

    High-performance, advanced nuclear power plant concepts have emerged with major emphasis on lower capital costs, inherent safety, and increased reliability. The materials problems posed by these concepts are discussed and how the scientists and technologists at ORNL plan to solve them is described

  1. Low Pressure Nuclear Thermal Rocket (LPNTR) concept

    International Nuclear Information System (INIS)

    Ramsthaler, J.H.

    1991-01-01

    A background and a description of the low pressure nuclear thermal system are presented. Performance, mission analysis, development, critical issues, and some conclusions are discussed. The following subject areas are covered: LPNTR's inherent advantages in critical NTR requirement; reactor trade studies; reference LPNTR; internal configuration and flow of preliminary LPNTR; particle bed fuel assembly; preliminary LPNTR neutronic study results; multiple LPNTR engine concept; tank and engine configuration for mission analysis; LPNTR reliability potential; LPNTR development program; and LPNTR program costs

  2. Advanced research workshop: nuclear materials safety

    International Nuclear Information System (INIS)

    Jardine, L J; Moshkov, M M.

    1999-01-01

    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

  3. Cross-cutting european thermal-hydraulics research for innovative nuclear systems

    International Nuclear Information System (INIS)

    Roelofs, F.; Class, A.; Cheng, X.; Meloni, P.; Van Tichelen, K.; Boudier, P.; Prasser, M.

    2010-01-01

    Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). This results in different micro- and macroscopic behavior of flow and heat transfer and requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulic issues are the subject of the 7. framework programme THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which runs from 2010 until 2014. This paper will describe the activities in this project which address the main identified thermal hydraulics issues for innovative nuclear systems. (authors)

  4. Thermal-hydraulics associated with nuclear education and research

    International Nuclear Information System (INIS)

    Yokobori, Seiichi

    2011-01-01

    This article was the rerecording of the author's lecture at the fourth 'Future Energy Forum' (aiming at improving nuclear safety and economics) held in December 2010. The lecture focused on (1) importance of thermal hydraulics associated with nuclear education and research (critical heat flux, two-phase flow and multiphase flow), (2) emerging trend of maintenance engineering (fluid induced vibration, flow accelerated corrosion and stress corrosion cracks), (3) fostering sensible nuclear engineer with common engineering sense, (4) balanced curriculum of basics and advanced research, (5) computerized simulation and fluid mechanics, (6) crucial point of thermo hydraulics education (viscosity, flux, steam and power generation), (7) safety education and human resources development (indispensable technologies such as defence in depth) and (8) topics of thermo hydraulics research (vortices of curbed pipes and visualization of two-phase flow). (T. Tanaka)

  5. Nuclear thermal propulsion engine cost trade studies

    International Nuclear Information System (INIS)

    Paschall, R.K.

    1993-01-01

    The NASA transportation strategy for the Mars Exploration architecture includes the use of nuclear thermal propulsion as the primary propulsion system for Mars transits. It is anticipated that the outgrowth of the NERVA/ROVER programs will be a nuclear thermal propulsion (NTP) system capable of providing the propulsion for missions to Mars. The specific impulse (Isp) for such a system is expected to be in the 870 s range. Trade studies were conducted to investigate whether or not it may be cost effective to invest in a higher performance (Isp>870 s) engine for nuclear thermal propulsion for missions to Mars. The basic cost trades revolved around the amount of mass that must be transported to low-earth orbit prior to each Mars flight and the cost to launch that mass. The mass required depended on the assumptions made for Mars missions scenarios including piloted/cargo flights, number of Mars missions, and transit time to Mars. Cost parameters included launch cost, program schedule for development and operations, and net discount rate. The results were very dependent on the assumptions that were made. Under some assumptions, higher performance engines showed cost savings in the billions of dollars; under other assumptions, the additional cost to develop higher performance engines was not justified

  6. Nuclear reactor vessel fuel thermal insulating barrier

    Science.gov (United States)

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  7. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)

    2015-12-01

    Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

  8. Multiphase Flow Dynamics 5 Nuclear Thermal Hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2012-01-01

    The present Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step...

  9. Multiphase flow dynamics 5 nuclear thermal hydraulics

    CERN Document Server

    Kolev, Nikolay Ivanov

    2015-01-01

    This Volume 5 of the successful book package "Multiphase Flow Dynamics" is devoted to nuclear thermal hydraulics which is a substantial part of nuclear reactor safety. It provides knowledge and mathematical tools for adequate description of the process of transferring the fission heat released in materials due to nuclear reactions into its environment. It step by step introduces into the heat release inside the fuel, temperature fields in the fuels, the "simple" boiling flow in a pipe described using ideas of different complexity like equilibrium, non equilibrium, homogeneity, non homogeneity. Then the "simple" three-fluid boiling flow in a pipe is described by gradually involving the mechanisms like entrainment and deposition, dynamic fragmentation, collisions, coalescence, turbulence. All heat transfer mechanisms are introduced gradually discussing their uncertainty. Different techniques are introduced like boundary layer treatments or integral methods. Comparisons with experimental data at each step demons...

  10. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    International Nuclear Information System (INIS)

    Heard, F.J.; Cramer, E.R.; Beaver, T.R.; Thurgood, M.J.

    1996-01-01

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy's Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses

  11. Shielding considerations for advanced space nuclear reactor systems

    International Nuclear Information System (INIS)

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

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO 2 ) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

  12. 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

  13. Local AREA networks in advanced nuclear reactors

    International Nuclear Information System (INIS)

    Bicknell, J.; Keats, A.B.

    1984-01-01

    The report assesses Local Area Network Communications with a view to their application in advanced nuclear reactor control and protection systems. Attention is focussed on commercially available techniques and systems for achieving the high reliability and availability required. A basis for evaluating network characteristics in terms of broadband or baseband type, medium, topology, node structure and access method is established. The reliability and availability of networks is then discussed. Several commercial networks are briefly assessed and a distinction made between general purpose networks and those suitable for process control. The communications requirements of nuclear reactor control and protection systems are compared with the facilities provided by current technology

  14. Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Weber, William J. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering; Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering

    2016-09-20

    This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effects of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.

  15. Nuclear thermal rockets using indigenous extraterrestrial propellants

    International Nuclear Information System (INIS)

    Zubrin, R.M.

    1990-01-01

    A preliminary examination of a concept for a Mars and outer solar system exploratory vehicle is presented. Propulsion is provided by utilizing a nuclear thermal reactor to heat a propellant volatile indigenous to the destination world to form a high thrust rocket exhaust. Candidate propellants, whose performance, materials compatibility, and ease of acquisition are examined and include carbon dioxide, water, methane, nitrogen, carbon monoxide, and argon. Ballistics and winged supersonic configurations are discussed. It is shown that the use of this method of propulsion potentially offers high payoff to a manned Mars mission. This is accomplished by sharply reducing the initial mission mass required in low earth orbit, and by providing Mars explorers with greatly enhanced mobility in traveling about the planet through the use of a vehicle that can refuel itself each time it lands. Thus, the nuclear landing craft is utilized in combination with a hydrogen-fueled nuclear-thermal interplanetary launch. By utilizing such a system in the outer solar system, a low level aerial reconnaissance of Titan combined with a multiple sample return from nearly every satellite of Saturn can be accomplished in a single launch of a Titan 4 or the Space Transportation System (STS). Similarly a multiple sample return from Callisto, Ganymede, and Europa can also be accomplished in one launch of a Titan 4 or the STS

  16. Applied mathematical methods in nuclear thermal hydraulics

    International Nuclear Information System (INIS)

    Ransom, V.H.; Trapp, J.A.

    1983-01-01

    Applied mathematical methods are used extensively in modeling of nuclear reactor thermal-hydraulic behavior. This application has required significant extension to the state-of-the-art. The problems encountered in modeling of two-phase fluid transients and the development of associated numerical solution methods are reviewed and quantified using results from a numerical study of an analogous linear system of differential equations. In particular, some possible approaches for formulating a well-posed numerical problem for an ill-posed differential model are investigated and discussed. The need for closer attention to numerical fidelity is indicated

  17. Nuclear thermal rocket engine operation and control

    International Nuclear Information System (INIS)

    Gunn, S.V.; Savoie, M.T.; Hundal, R.

    1993-06-01

    The operation of a typical Rover/Nerva-derived nuclear thermal rocket (NTR) engine is characterized and the control requirements of the NTR are defined. A rationale for the selection of a candidate diverse redundant NTR engine control system is presented and the projected component operating requirements are related to the state of the art of candidate components and subsystems. The projected operational capabilities of the candidate system are delineated for the startup, full-thrust, shutdown, and decay heat removal phases of the engine operation. 9 refs

  18. 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.

  19. 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.

  20. Advanced nuclear reactor public opinion project

    International Nuclear Information System (INIS)

    Benson, B.

    1991-01-01

    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

  1. Equipment system for advanced nuclear fuel development

    International Nuclear Information System (INIS)

    Kwon, Hyuk Il; Ji, C. G.; Bae, S. O.

    2002-11-01

    The purpose of the settlement of equipment system for nuclear Fuel Technology Development Facility(FTDF) is to build a seismic designed facility that can accommodate handling of nuclear materials including <20% enriched Uranium and produce HANARO fuel commercially, and also to establish the advanced common research equipment essential for the research on advanced fuel development. For this purpose, this research works were performed for the settlement of radiation protection system and facility special equipment for the FTDF, and the advanced common research equipment for the fuel fabrication and research. As a result, 11 kinds of radiation protection systems such as criticality detection and alarm system, 5 kinds of facility special equipment such as environmental pollution protection system and 5 kinds of common research equipment such as electron-beam welding machine were established. By the settlement of exclusive domestic facility for the research of advanced fuel, the fabrication and supply of HANARO fuel is possible and also can export KAERI-invented centrifugal dispersion fuel materials and its technology to the nations having research reactors in operation. For the future, the utilization of the facility will be expanded to universities, industries and other research institutes

  2. On the safety performance of the advanced nuclear energy systems

    International Nuclear Information System (INIS)

    Li Shounan

    1999-01-01

    Some features on the safety performances of the Advanced Nuclear Energy Systems are discussed. The advantages and some peculiar problems on the safety of Advanced Nuclear Energy Systems with subcritical nuclear reactor driven by external neutron sources are also pointed out in comparison with conventional nuclear reactors

  3. Advanced physical protection systems for nuclear materials

    International Nuclear Information System (INIS)

    Jones, O.E.

    1975-10-01

    Because of the increasing incidence of terrorism, there is growing concern that nuclear materials and facilities need improved physical protection against theft, diversion, or sabotage. Physical protection systems for facilities or transportation which have balanced effectiveness include information systems, access denial systems, adequate and timely response, recovery capability, and use denial methods for despoiling special nuclear materials (SNM). The role of these elements in reducing societal risk is described; however, it is noted that, similar to nuclear war, the absolute risks of nuclear theft and sabotage are basically unquantifiable. Sandia Laboratories has a major Energy Research and Development Administration (ERDA) role in developing advanced physical protection systems for improving the security of both SNM and facilities. These activities are surveyed. A computer simulation model is being developed to assess the cost-effectiveness of alternative physical protection systems under various levels of threat. Improved physical protection equipment such as perimeter and interior alarms, secure portals, and fixed and remotely-activated barriers is being developed and tested. In addition, complete prototype protection systems are being developed for representative nuclear facilities. An example is shown for a plutonium storage vault. The ERDA safe-secure transportation system for highway shipments of all significant quantities of government-owned SNM is described. Adversary simulation as a tool for testing and evaluating physical protection systems is discussed. A list of measures is given for assessing overall physical protection system performance. (auth)

  4. Advanced physical protection systems for nuclear materials

    International Nuclear Information System (INIS)

    Jones, O.E.

    1976-01-01

    Because of the increasing incidence of terrorism, there is growing concern that nuclear materials and facilities need improved physical protection against theft, diversion, or sabotage. Physical protection systems for facilities or transportation which have balanced effectiveness include information systems, access denial systems, adequate and timely response, recovery capability, and use denial methods for despoiling special nuclear materials (SNM). The role of these elements in reducing societal risk is described; however, it is noted that, similar to nuclear war, the absolute risks of nuclear theft and sabotage are basically unquantifiable. Sandia Laboratories has a major US Energy Research and Development Administration (ERDA) role in developing advanced physical protection systems for improving the security of both SNM and facilities. These activities are surveyed in this paper. A computer simulation model is being developed to assess the cost-effectiveness of alternative physical protection systems under various levels of threat. Improved physical protection equipment such as perimeter and interior alarms, secure portals, and fixed and remotely activated barriers is being developed and tested. In addition, complete prototype protection systems are being developed for representative nuclear facilities. An example is shown for a plutonium storage vault. The ERDA safe-secure transportation system for highway shipments of all significant quantities of government-owned SNM is described. Adversary simulation as a tool for testing and evaluating physical protection systems is discussed. Finally, a list of measures is given for assessing overall physical protection system performance. (author)

  5. Status of advanced nuclear reactor development in Korea

    International Nuclear Information System (INIS)

    Kim, H.R.; Kim, K.K.; Kim, Y.W.; Joo, H.K.

    2014-01-01

    The Korean nuclear industry is facing new challenges to solve the spent fuel storage problem and meet the needs to diversify the application areas of nuclear energy. In order to provide solutions to these challenges, the Korea Atomic Energy Research Institute (KAERI) has been developing advanced nuclear reactors including a Sodium-cooled Fast Reactor, Very High Temperature Gas cooled Reactor (VHTR), and System-integrated Modular Advanced Reactor (SMART) with substantially improved safety, economics, and environment-friendly features. A fast reactor system is one of the most promising options for a reduction of radioactive wastes. The long-term plan for Advanced SFR development in conjunction with the pyro-process was authorized by the Korean Atomic Energy Commission in 2008. The development milestone includes specific design approval of a prototype SFR by 2020, and the construction of a prototype SFR by 2028. KAERI has been carrying out the preliminary design of a 150MWe SFR prototype plant system since 2012. The development of advanced SFR technologies and the basic key technologies necessary for the prototype SFR are also being carried out. By virtue of high-temperature heat, a VHTR has diverse applications including hydrogen production. KAERI launched a nuclear hydrogen project using a VHTR in 2006, which focused on four basic technologies: the development of design tools, very high-temperature experimental technology, TRISO fuel fabrication, and Sulfur-iodine thermo-chemical hydrogen production technology. The technology development project will be continued until 2017. A conceptual reactor design study was started in 2012 as collaboration between industry and government to enhance the early-launching of the nuclear hydrogen development and demonstration (NHDD) project. The goal of the NHDD project is to design and build a nuclear hydrogen demonstration system by 2030. KAERI has developed SMART which is a small-sized advanced integral reactor with a rated

  6. Proceedings of the third nuclear thermal hydraulics meeting

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    This book contains the proceedings of the Thermal Hydraulics Division of the American Nuclear Society. The papers presented include: Simulator qualification using engineering codes and Development of thermal hydraulic analysis capabilities for Oyster Creek

  7. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Baigorria, Carlos.

    1983-12-01

    Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor) [es

  8. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

  9. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  10. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

  11. Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

    International Nuclear Information System (INIS)

    2004-01-01

    More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

  12. Thermal Model Predictions of Advanced Stirling Radioisotope Generator Performance

    Science.gov (United States)

    Wang, Xiao-Yen J.; Fabanich, William Anthony; Schmitz, Paul C.

    2014-01-01

    This presentation describes the capabilities of three-dimensional thermal power model of advanced stirling radioisotope generator (ASRG). The performance of the ASRG is presented for different scenario, such as Venus flyby with or without the auxiliary cooling system.

  13. Advanced nuclear fuel cycles activities in IAEA

    International Nuclear Information System (INIS)

    Nawada, H.P.; Ganguly, C.

    2007-01-01

    Full text of publication follows. Of late several developments in reprocessing areas along with advances in fuel design and robotics have led to immense interest in partitioning and transmutation (P and T). The R and D efforts in the P and T area are being paid increased attention as potential answers to ever-growing issues threatening sustainability, environmental protection and non-proliferation. Any fuel cycle studies that integrate partitioning and transmutation are also known as ''advanced fuel cycles'' (AFC), that could incinerate plutonium and minor actinide (MA) elements (namely Am, Np, Cm, etc.) which are the main contributors to long-term radiotoxicity. The R and D efforts in developing these innovative fuel cycles as well as reactors are being co-ordinated by international initiatives such as Innovative Nuclear Power Reactors and Fuel Cycles (INPRO), the Generation IV International Forum (GIF) and the Global Nuclear Energy Partnership (GENP). For these advanced nuclear fuel cycle schemes to take shape, the development of liquid-metal-cooled reactor fuel cycles would be the most essential step for implementation of P and T. Some member states are also evaluating other concepts involving the use of thorium fuel cycle or inert-matrix fuel or coated particle fuel. Advanced fuel cycle involving novel partitioning methods such as pyrochemical separation methods to recover the transuranic elements are being developed by some member states which would form a critical stage of P and T. However, methods that can achieve a very high reduction (>99.5%) of MA and long-lived fission products in the waste streams after partitioning must be achieved to realize the goal of an improved protection of the environment. In addition, the development of MA-based fuel is also an essential and crucial step for transmutation of these transuranic elements. The presentation intends to describe progress of the IAEA activities encompassing the following subject-areas: minimization of

  14. Advances in solar thermal energy in Uruguay

    International Nuclear Information System (INIS)

    Franco Noceto, P.

    2012-01-01

    This article is about the law 18585 which declared de solar thermal energy as national interest. This law establishes the obligation to incorporate solar heating systems in health care centers, hotels and sports clubs.

  15. Oxygen Containment System Options for Nuclear Thermal Propulsion Testing

    Data.gov (United States)

    National Aeronautics and Space Administration — All nuclear thermal propulsion (NTP) ground testing conducted in the 1950s and 1960s during the ROVER/(Nuclear Engine Rocket Vehicle Application (NERVA) program...

  16. Nuclear vapor thermal reactor propulsion technology

    International Nuclear Information System (INIS)

    Maya, I.; Diaz, N.J.; Dugan, E.T.; Watanabe, Y.; McClanahan, J.A.; Wen-Hsiung Tu; Carman, R.L.

    1993-01-01

    The conceptual design of a nuclear rocket based on the vapor core reactor is presented. The Nuclear Vapor Thermal Rocket (NVTR) offers the potential for a specific impulse of 1000 to 1200 s at thrust-to-weight ratios of 1 to 2. The design is based on NERVA geometry and systems with the solid fuel replaced by uranium tetrafluoride (UF 4 ) vapor. The closed-loop core does not rely on hydrodynamic confinement of the fuel. The hydrogen propellant is separated from the UF 4 fuel gas by graphite structure. The hydrogen is maintained at high pressure (∼100 atm), and exits the core at 3,100 K to 3,500 K. Zirconium carbide and hafnium carbide coatings are used to protect the hot graphite from the hydrogen. The core is surrounded by beryllium oxide reflector. The nuclear reactor core has been integrated into a 75 klb engine design using an expander cycle and dual turbopumps. The NVTR offers the potential for an incremental technology development pathway to high performance gas core reactors. Since the fuel is readily available, it also offers advantages in the initial cost of development, as it will not require major expenditures for fuel development

  17. Advanced Thermophotovoltaic Devices for Space Nuclear Power Systems

    International Nuclear Information System (INIS)

    Wernsman, Bernard; Mahorter, Robert G.; Siergiej, Richard; Link, Samuel D.; Wehrer, Rebecca J.; Belanger, Sean J.; Fourspring, Patrick; Murray, Susan; Newman, Fred; Taylor, Dan; Rahmlow, Tom

    2005-01-01

    Advanced thermophotovoltaic (TPV) modules capable of producing > 0.3 W/cm2 at an efficiency > 22% while operating at a converter radiator and module temperature of 1228 K and 325 K, respectively, have been made. These advanced TPV modules are projected to produce > 0.9 W/cm2 at an efficiency > 24% while operating at a converter radiator and module temperature of 1373 K and 325 K, respectively. Radioisotope and nuclear (fission) powered space systems utilizing these advanced TPV modules have been evaluated. For a 100 We radioisotope TPV system, systems utilizing as low as 2 general purpose heat source (GPHS) units are feasible, where the specific power for the 2 and 3 GPHS unit systems operating in a 200 K environment is as large as ∼ 16 We/kg and ∼ 14 We/kg, respectively. For a 100 kWe nuclear powered (as was entertained for the thermoelectric SP-100 program) TPV system, the minimum system radiator area and mass is ∼ 640 m2 and ∼ 1150 kg, respectively, for a converter radiator, system radiator and environment temperature of 1373 K, 435 K and 200 K, respectively. Also, for a converter radiator temperature of 1373 K, the converter volume and mass remains less than 0.36 m3 and 640 kg, respectively. Thus, the minimum system radiator + converter (reactor and shield not included) specific mass is ∼ 16 kg/kWe for a converter radiator, system radiator and environment temperature of 1373 K, 425 K and 200 K, respectively. Under this operating condition, the reactor thermal rating is ∼ 1110 kWt. Due to the large radiator area, the added complexity and mission risk needs to be weighed against reducing the reactor thermal rating to determine the feasibility of using TPV for space nuclear (fission) power systems

  18. Safety related terms for advanced nuclear plants

    International Nuclear Information System (INIS)

    1995-12-01

    The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

  19. Safety related terms for advanced nuclear plants

    International Nuclear Information System (INIS)

    1991-09-01

    The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

  20. Advanced Testing Method for Ground Thermal Conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiaobing [ORNL; Clemenzi, Rick [Geothermal Design Center Inc.; Liu, Su [University of Tennessee (UT)

    2017-04-01

    A new method is developed that can quickly and more accurately determine the effective ground thermal conductivity (GTC) based on thermal response test (TRT) results. Ground thermal conductivity is an important parameter for sizing ground heat exchangers (GHEXs) used by geothermal heat pump systems. The conventional GTC test method usually requires a TRT for 48 hours with a very stable electric power supply throughout the entire test. In contrast, the new method reduces the required test time by 40%–60% or more, and it can determine GTC even with an unstable or intermittent power supply. Consequently, it can significantly reduce the cost of GTC testing and increase its use, which will enable optimal design of geothermal heat pump systems. Further, this new method provides more information about the thermal properties of the GHEX and the ground than previous techniques. It can verify the installation quality of GHEXs and has the potential, if developed, to characterize the heterogeneous thermal properties of the ground formation surrounding the GHEXs.

  1. Licensed bases management for advanced nuclear plants

    International Nuclear Information System (INIS)

    O'Connell, J.; Rumble, E.; Rodwell, E.

    2001-01-01

    Prospective Advanced Nuclear Plant (ANP) owners must have high confidence that the integrity of the licensed bases (LB) of a plant will be effectively maintained over its life cycle. Currently, licensing engineers use text retrieval systems, database managers, and checklists to access, update, and maintain vast and disparate licensing information libraries. This paper describes the demonstration of a ''twin-engine'' approach that integrates a program from the emerging class of concept searching tools with a modern Product Data Management System (PDMS) to enhance the management of LB information for an example ANP design. (author)

  2. Licensed bases management for advanced nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    O' Connell, J [Duke Engineering and Services, Marlborough, MA (United States); Rumble, E; Rodwell, E [EPRI, Palo Alto, CA (United States)

    2001-07-01

    Prospective Advanced Nuclear Plant (ANP) owners must have high confidence that the integrity of the licensed bases (LB) of a plant will be effectively maintained over its life cycle. Currently, licensing engineers use text retrieval systems, database managers, and checklists to access, update, and maintain vast and disparate licensing information libraries. This paper describes the demonstration of a ''twin-engine'' approach that integrates a program from the emerging class of concept searching tools with a modern Product Data Management System (PDMS) to enhance the management of LB information for an example ANP design. (author)

  3. Licensed bases management for advanced nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    O' Connell, J. [Duke Engineering and Services, Marlborough, MA (United States); Rumble, E.; Rodwell, E. [EPRI, Palo Alto, CA (United States)

    2001-07-01

    Prospective Advanced Nuclear Plant (ANP) owners must have high confidence that the integrity of the licensed bases (LB) of a plant will be effectively maintained over its life cycle. Currently, licensing engineers use text retrieval systems, database managers, and checklists to access, update, and maintain vast and disparate licensing information libraries. This paper describes the demonstration of a ''twin-engine'' approach that integrates a program from the emerging class of concept searching tools with a modern Product Data Management System (PDMS) to enhance the management of LB information for an example ANP design. (author)

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

    Directory of Open Access Journals (Sweden)

    Wang Qiu-kuan

    2012-01-01

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

  5. Advanced pyrochemical technologies for minimizing nuclear waste

    International Nuclear Information System (INIS)

    Bronson, M.C.; Dodson, K.E.; Riley, D.C.

    1994-01-01

    The Department of Energy (DOE) is seeking to reduce the size of the current nuclear weapons complex and consequently minimize operating costs. To meet this DOE objective, the national laboratories have been asked to develop advanced technologies that take uranium and plutonium, from retired weapons and prepare it for new weapons, long-term storage, and/or final disposition. Current pyrochemical processes generate residue salts and ceramic wastes that require aqueous processing to remove and recover the actinides. However, the aqueous treatment of these residues generates an estimated 100 liters of acidic transuranic (TRU) waste per kilogram of plutonium in the residue. Lawrence Livermore National Laboratory (LLNL) is developing pyrochemical techniques to eliminate, minimize, or more efficiently treat these residue streams. This paper will present technologies being developed at LLNL on advanced materials for actinide containment, reactors that minimize residues, and pyrochemical processes that remove actinides from waste salts

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

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

  7. Advanced Fuel Cell System Thermal Management for NASA Exploration Missions

    Science.gov (United States)

    Burke, Kenneth A.

    2009-01-01

    The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA exploration program. An analysis of a state-of-the-art fuel cell cooling systems was done to benchmark the portion of a fuel cell system s mass that is dedicated to thermal management. Additional analysis was done to determine the key performance targets of the advanced passive thermal management technology that would substantially reduce fuel cell system mass.

  8. Thermal effects and their compensation in Advanced Virgo

    International Nuclear Information System (INIS)

    Rocchi, A; Coccia, E; Fafone, V; Malvezzi, V; Minenkov, Y; Sperandio, L

    2012-01-01

    Thermal effects in the test masses of the gravitational waves interferometric detectors may result in a strong limitation to their operation and sensitivity. Already in initial LIGO and Virgo, these effects have been observed and required the installation of dedicated compensation systems. Based on CO 2 laser projectors, the thermal compensators heat the peripheral of the input test masses to reduce the lensing effect. In advanced detectors, the power circulating in the interferometer will increase, thus making thermal effects more relevant. In this paper, the concept of the compensation system for Advanced Virgo is described.

  9. Thermally-insulating layer for nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    The thermally-insulating layer has been designed both for insulating surfaces within the core of a nuclear reactor and transmitting loads such as the core-weight. Said layer comprises a layer of bricks and a layer of tiles with smaller clearance between the tiles than between the bricks, the latter having a reduced cross-section against the tiles so as to be surrounded by relatively large interconnected ducts forming a continuous chamber behind the tile-layer in order to induce a substantial decreases in the transverse flow of the reactor-core coolant. The core preferably comprises hexagonal columns supported by rhomb-shaped plates, with channels distributed so as to mix the coolant of twelve columns. The plates are separated from support-tiles by means of pillars [fr

  10. NASA's nuclear thermal propulsion technology project

    International Nuclear Information System (INIS)

    Peecook, K.M.; Stone, J.R.

    1992-07-01

    The nonnuclear subsystem technologies required for incorporating nuclear thermal propulsion (NTP) into space-exploration missions are discussed. Of particular interest to planned missions are such technologies as materials, instrumentation and controls, turbomachinery, CFD modeling, nozzle extension designs and models, and analyses of exhaust plumes. NASA studies are described and/or proposed for refractory metals and alloys, robotic NTP controls, and turbopump materials candidates. Alternative nozzle concepts such as aerospikes and truncated plugs are proposed, and numerical simulations are set forth for studying heavy molecules and the backstreaming of highly reactive free-radical hydrogen in the exhaust plume. The critical technologies described in the paper are central to the development of NTP, and NTP has the potential to facilitate a range of space exploration activities. 3 refs

  11. System model development for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Walton, J.T.; Perkins, K.R.; Buksa, J.J.; Worley, B.A.; Dobranich, D.

    1992-01-01

    A critical enabling technology in the evolutionary development of nuclear thermal propulsion (NTP) is the ability to predict the system performance under a variety of operating conditions. Since October 1991, US (DOE), (DOD) and NASA have initiated critical technology development efforts for NTP systems to be used on Space Exploration Initiative (SEI) missions to the Moon and Mars. This paper presents the strategy and progress of an interagency NASA/DOE/DOD team for NTP system modeling. It is the intent of the interagency team to develop several levels of computer programs to simulate various NTP systems. An interagency team was formed for this task to use the best capabilities available and to assure appropriate peer review. The vision and strategy of the interagency team for developing NTP system models will be discussed in this paper. A review of the progress on the Level 1 interagency model is also presented

  12. Thermal analyses of spent nuclear fuel repository

    International Nuclear Information System (INIS)

    Ikonen, K.

    2003-06-01

    This report contains the temperature dimensioning of the KBS-3V type 1- or 2-panel repository based on the rock properties measured from the Olkiluoto investigations. The report describes first the development of a calculation methodology for the thermal analysis of a repository for nuclear fuel. The disposed canisters produce residual heat due to decay (or disintegration) of radioactive products. The decay heat is conducted to surrounding rock mass. The methods were applied to determine the effect of different parameters on the highest canister temperature and to support the planning, dimensioning and operation of the repository. The thermal diffusivity of the rock is low and the heat released from the canisters is spread into the surrounding rock volume quite slowly causing thermal gradient in the rock close to canisters and the canister temperature is increased remarkably. The maximum temperature on the canister surface is limited to the design temperature of +100 deg C. However, due to uncertainties in thermal analysis parameters (like scattering in rock conductivity) the allowable calculated maximum canister temperature is set to 90 deg C causing a safety margin of 10 deg C. The allowable temperature is controlled by the spacing between adjacent canisters, adjacent tunnels and the distance between separate panels of the repository and the pre-cooling time affecting power of the canisters. Because of the fact that the disposal operation takes several decades, the moment of disposal of an individual canister in addition to the location has an influence on the maximum temperature in the canister. Also, a second disposal panel in the repository has a thermal interaction with the other panel. This interaction is expressed after a few decades at the strongest. It became apparent that the temperature of canister surfaces can be determined by analytic line heat source model much more efficiently than by numerical analysis, if the analytic model is first verified and

  13. Next generation advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    Turgut, M. H.

    2009-01-01

    Growing energy demand by technological developments and the increase of the world population and gradually diminishing energy resources made nuclear power an indispensable option. The renewable energy sources like solar, wind and geothermal may be suited to meet some local needs. Environment friendly nuclear energy which is a suitable solution to large scale demands tends to develop highly economical, advanced next generation reactors by incorporating technological developments and years of operating experience. The enhancement of safety and reliability, facilitation of maintainability, impeccable compatibility with the environment are the goals of the new generation reactors. The protection of the investment and property is considered as well as the protection of the environment and mankind. They became economically attractive compared to fossil-fired units by the use of standard designs, replacing some active systems by passive, reducing construction time and increasing the operation lifetime. The evolutionary designs were introduced at first by ameliorating the conventional plants, than revolutionary systems which are denoted as generation IV were verged to meet future needs. The investigations on the advanced, proliferation resistant fuel cycle technologies were initiated to minimize the radioactive waste burden by using new generation fast reactors and ADS transmuters.

  14. Thermochemistry of nuclear fuels in advanced reactors

    International Nuclear Information System (INIS)

    Agarwal, Renu

    2015-01-01

    The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

  15. Ultrahigh Specific Impulse Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Anne Charmeau; Brandon Cunningham; Samim Anghaie

    2009-02-09

    Research on nuclear thermal propulsion systems (NTP) have been in forefront of the space nuclear power and propulsion due to their design simplicity and their promise for providing very high thrust at reasonably high specific impulse. During NERVA-ROVER program in late 1950's till early 1970's, the United States developed and ground tested about 18 NTP systems without ever deploying them into space. The NERVA-ROVER program included development and testing of NTP systems with very high thrust (~250,000 lbf) and relatively high specific impulse (~850 s). High thrust to weight ratio in NTP systems is an indicator of high acceleration that could be achieved with these systems. The specific impulse in the lowest mass propellant, hydrogen, is a function of square root of absolute temperature in the NTP thrust chamber. Therefor optimizing design performance of NTP systems would require achieving the highest possible hydrogen temperature at reasonably high thrust to weight ratio. High hydrogen exit temperature produces high specific impulse that is a diret measure of propellant usage efficiency.

  16. Nuclear Thermal Rocket Simulation in NPSS

    Science.gov (United States)

    Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

    2013-01-01

    Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

  17. Nuclear techniques in the development of advanced ceramic technologies

    International Nuclear Information System (INIS)

    Axe, J.D.; Hewat, A.W.; Maier, J.; Margaca, F.M.A.; Rauch, H.

    1999-01-01

    The importance of research, development and application of advanced materials is well understood by all developed and most developing countries. Amongst advanced materials, ceramics play a prominent role due to their specific chemical and physical properties. According to performance and importance, advanced ceramics can be classified as structural ceramics (mechanical function) and the so-called functional ceramics. In the latter class of materials, special electrical, chemical, thermal, magnetic and optical properties are of interest. The most valuable materials are multifunctional, for example, when structural ceramics combine beneficial mechanical properties with thermal and chemical sensitivity. Multifunctionality is characteristic of many composite materials (organic/inorganic composite). Additionally, properties of material can be changed by reducing its dimension (thin films, nanocrystalline ceramics). Nuclear techniques, found important applications in research and development of advanced ceramics. The use of neutron techniques has increased dramatically in recent years due to the development of advanced neutron sources, instrumentation and improved data analysis. Typical neutron techniques are neutron diffraction, neutron radiography, small angle neutron scattering and very small angle neutron scattering. Neutrons can penetrate deeply into most materials thus sampling their bulk properties. In determination of the crystal structure of HTSC, YBa 2 Cu 2 O 7 , XRD located the heavy metal atoms, but failed in finding many of the oxygen atoms, while the neutron diffraction located all atoms equally well in the crystal structure. Neutron diffraction is also unique for the determination of the magnetic structure of materials since the neutrons themselves have a magnetic moment. Application of small angle neutron scattering for the determination of the size of hydrocarbon aggregates within the zeolite channels is illustrated. (author)

  18. Nuclear Thermal Rocket (NTR) Development Risk Communication

    Science.gov (United States)

    Kim, Tony

    2014-01-01

    There are clear advantages of development of a Nuclear Thermal Rocket (NTR) for a crewed mission to Mars. NTR for in-space propulsion enables more ambitious space missions by providing high thrust at high specific impulse (approximately 900 sec) that is 2 times the best theoretical performance possible for chemical rockets. Missions can be optimized for maximum payload capability to take more payload with reduced total mass to orbit; saving cost on reduction of the number of launch vehicles needed. Or missions can be optimized to minimize trip time significantly to reduce the deep space radiation exposure to the crew. NTR propulsion technology is a game changer for space exploration. However, "NUCLEAR" is a word that is feared and vilified by some groups and the hostility towards development of any nuclear systems can meet great opposition by the public as well as from national leaders and people in authority. Communication of nuclear safety will be critical to the success of the development of the NTR. Why is there a fear of nuclear? A bomb that can level a city is a scary weapon. The first and only times the Nuclear Bomb was used in a war was on Hiroshima and Nagasaki during World War 2. The "Little Boy" atomic bomb was dropped on Hiroshima on August 6, 1945 and the "Fat Man" on Nagasaki 3 days later on August 9th. Within the first 4 months of bombings, 90- 166 thousand people died in Hiroshima and 60-80 thousand died in Nagasaki. It is important to note for comparison that over 500 thousand people died and 5 million made homeless due to strategic bombing (approximately 150 thousand tons) of Japanese cities and war assets with conventional non-nuclear weapons between 1942- 1945. A major bombing campaign of "firebombing" of Tokyo called "Operation Meetinghouse" on March 9 and 10 consisting of 334 B-29's dropped approximately1,700 tons of bombs around 16 square mile area and over 100 thousand people have been estimated to have died. The declaration of death is very

  19. A new advanced safe nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, Farhang

    1999-01-01

    The reactor design is based on fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident condition. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed in order to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another is to modify the shape of the reactor core to produce a non-fluctuating bed and consequently guarantee the dynamic stability of the reactor. The mixing of Tantalum in the fuel is also proposed as an additional inhibition to power excursion. The spent fuel pellets may not be considered nuclear waste since they are in the shape and size that can easily be used as a a radioactive source for food irradiation and industrial applications. The reactor can easily operate with any desired spectrum by varying the porosity in order to be a plutonium burner or utilize a thorium fuel cycle. (author)

  20. Fabrication of High Temperature Cermet Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Panda, Binayak; Shah, Sandeep

    2005-01-01

    Processing techniques are being developed to fabricate refractory metal and ceramic cermet materials for Nuclear Thermal Propulsion (NTP). Significant advances have been made in the area of high-temperature cermet fuel processing since RoverNERVA. Cermet materials offer several advantages such as retention of fission products and fuels, thermal shock resistance, hydrogen compatibility, high conductivity, and high strength. Recent NASA h d e d research has demonstrated the net shape fabrication of W-Re-HfC and other refractory metal and ceramic components that are similar to UN/W-Re cermet fuels. This effort is focused on basic research and characterization to identify the most promising compositions and processing techniques. A particular emphasis is being placed on low cost processes to fabricate near net shape parts of practical size. Several processing methods including Vacuum Plasma Spray (VPS) and conventional PM processes are being evaluated to fabricate material property samples and components. Surrogate W-Re/ZrN cermet fuel materials are being used to develop processing techniques for both coated and uncoated ceramic particles. After process optimization, depleted uranium-based cermets will be fabricated and tested to evaluate mechanical, thermal, and hot H2 erosion properties. This paper provides details on the current results of the project.

  1. Radioactive waste management and advanced nuclear fuel cycle technologies

    International Nuclear Information System (INIS)

    2007-01-01

    In 2007 ENEA's Department of Nuclear Fusion and Fission, and Related Technologies acted according to national policy and the role assigned to ENEA FPN by Law 257/2003 regarding radioactive waste management and advanced nuclear fuel cycle technologies

  2. Thermal efficiency improvements - an imperative for nuclear generating stations

    International Nuclear Information System (INIS)

    Hassanien, S.; Rouse, S.

    1997-01-01

    A one and a half percent thermal performance improvement of Ontario Hydro's operating nuclear units (Bruce B, Pickering B, and Darlington) means almost 980 GWh are available to the transmission system (assuming an 80% capacity factor). This is equivalent to the energy consumption of 34,000 electrically-heated homes in Ontario, and worth more than $39 million in revenue to Ontario Hydro Nuclear Generation. Improving nuclear plant thermal efficiency improves profitability (more GWh per unit of fuel) and competitiveness (cost of unit energy), and reduces environmental impact (less spent fuel and nuclear waste). Thermal performance will naturally decrease due to the age of the units unless corrective action is taken. Most Ontario Hydro nuclear units are ten to twenty years old. Some common causes for loss of thermal efficiency are: fouling and tube plugging of steam generators, condensers, and heat exchangers; steam leaks in the condenser due to valve wear, steam trap and drain leaks; deposition, pitting, cracking, corrosion, etc., of turbine blades; inadequate feedwater metering resulting from corrosion and deposition. This paper stresses the importance of improving the nuclear units' thermal efficiency. Ontario Hydro Nuclear has demonstrated energy savings results are achievable and affordable. Between 1994 and 1996, Nuclear reduced its energy use and improved thermal efficiency by over 430,000 MWh. Efficiency improvement is not automatic - strategies are needed to be effective. This paper suggests practical strategies to systematically improve thermal efficiency. (author)

  3. Space nuclear thermal propulsion test facilities accommodation at INEL

    International Nuclear Information System (INIS)

    Hill, T.J.; Reed, W.C.; Welland, H.J.

    1993-01-01

    The U.S. Air Force (USAF) has proposed to develop the technology and demonstrate the feasibility of a particle bed reactor (PBR) propulsion system that could be used to power an advanced upper stage rocket engine. The U.S. Department of Energy (DOE) is cooperating with the USAF in that it would host the test facility if the USAF decides to proceed with the technology demonstration. Two DOE locations have been proposed for testing the PBR technology, a new test facility at the Nevada Test Site, or the modification and use of an existing facility at the Idaho National Engineering Laboratory. The preliminary evaluations performed at the INEL to support the PBR technology testing has been completed. Additional evaluations to scope the required changes or upgrade needed to make the proposed USAF PBR test facility meet the requirements for testing Space Exploration Initiative (SEI) nuclear thermal propulsion engines are underway

  4. Space nuclear thermal propulsion test facilities accommodation at INEL

    Science.gov (United States)

    Hill, Thomas J.; Reed, William C.; Welland, Henry J.

    1993-01-01

    The U.S. Air Force (USAF) has proposed to develop the technology and demonstrate the feasibility of a particle bed reactor (PBR) propulsion system that could be used to power an advanced upper stage rocket engine. The U.S. Department of Energy (DOE) is cooperating with the USAF in that it would host the test facility if the USAF decides to proceed with the technology demonstration. Two DOE locations have been proposed for testing the PBR technology, a new test facility at the Nevada Test Site, or the modification and use of an existing facility at the Idaho National Engineering Laboratory. The preliminary evaluations performed at the INEL to support the PBR technology testing has been completed. Additional evaluations to scope the required changes or upgrade needed to make the proposed USAF PBR test facility meet the requirements for testing Space Exploration Initiative (SEI) nuclear thermal propulsion engines are underway.

  5. Advanced nuclear plants: Meeting the economic challenge

    International Nuclear Information System (INIS)

    Redding, J.R.; MacGregor, P.R.

    1993-01-01

    As the end of the century draws closer, utilities all over the world face a daunting challenge. Forecasts, such as those prepared by General Electric show a dramatic increase in the need for additional electrical generating capacity (>1000 GW for the world from 1996 through 2006). In the case of the United States, such increases are almost one-third of existing capacity. Furthermore, this capacity must be environmentally benign to satisfy mounting concerns over the environmental impact of burning fossil fuels. This paper discusses economic analyses, similar to least-cost option studies, conducted by General Electric to determine the cost levels that new, advanced nuclear plants must meet to compete with other available generating options

  6. Westinghouse AP600 advanced nuclear plant design

    International Nuclear Information System (INIS)

    Gangloff, W.

    1999-01-01

    As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

  7. Nuclear integrated database and design advancement system

    International Nuclear Information System (INIS)

    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

  8. Advanced materials for thermal management of electronic packaging

    CERN Document Server

    Tong, Xingcun Colin

    2011-01-01

    The need for advanced thermal management materials in electronic packaging has been widely recognized as thermal challenges become barriers to the electronic industry's ability to provide continued improvements in device and system performance. With increased performance requirements for smaller, more capable, and more efficient electronic power devices, systems ranging from active electronically scanned radar arrays to web servers all require components that can dissipate heat efficiently. This requires that the materials have high capability of dissipating heat and maintaining compatibility

  9. Research on technology of evaluating thermal property data of nuclear power materials

    International Nuclear Information System (INIS)

    Imai, Hidetaka; Baba, Tetsuya; Matsumoto, Tsuyoshi; Kishimoto, Isao; Taketoshi, Naoyuki; Arai, Teruo

    1997-01-01

    For the materials of first wall and diverter of nuclear fusion reactor, in order to withstand steady and unsteady high heat flux load, excellent thermal characteristics are required. It is strongly demanded to measure such thermal property values as heat conductivity, heat diffusivity, specific heat capacity, emissivity and so using small test pieces up to higher than 2000degC. As the materials of nuclear reactors are subjected to neutron irradiation, in order to secure the long term reliability of the materials, it is very important to establish the techniques for forecasting the change of the thermal property values due to irradiation effect. Also the establishment of the techniques for estimating the thermal property values of new materials like low radioactivation material is important. In National Research Laboratory of Metrology, the research on the advancement of the measuring technology for high temperature thermal properties has resulted in the considerably successful development of such technologies. In this research, the rapid measurement of thermal property values up to superhigh temperature with highest accuracy, the making of thermal property data set of high level, the analysis and evaluation of the correlation of material characters and thermal property values, and the development of the basic techniques for estimating the thermal property values of solid materials are aimed at and advanced. These are explained. (K.I.)

  10. Engineering thermal engine rocket adventurer for space nuclear application

    International Nuclear Information System (INIS)

    Nam, Seung H.; Suh, Kune Y.; Kang, Seong G.

    2008-01-01

    The conceptual design for the first-of-a-kind engineering of Thermal Engine Rocket Adventure (TERA) is described. TERA comprising the Battery Omnibus Reactor Integral System (BORIS) as the heat resource and the Space Propulsion Reactor Integral System (SPRIS) as the propulsion system, is one of the advanced Nuclear Thermal Rocket (NTR) engine utilizing hydrogen (H 2 ) propellant being developed at present time. BORIS in this application is an open cycle high temperature gas cooled reactor that has eighteen fuel elements for propulsion and one fuel element for electricity generation and propellant pumping. Each fuel element for propulsion has its own small nozzle. The nineteen fuel elements are arranged into hexagonal prism shape in the core and surrounded by outer Be reflector. The TERA maximum power is 1,000 MW th , specific impulse 1,000 s, thrust 250,000 N, and the total mass is 550 kg including the reactor, turbo pump and auxiliaries. Each fuel element comprises the fuel assembly, moderators, pressure tube and small nozzle. The TERA fuel assembly is fabricated of 93% enriched 1.5 mm (U, Zr, Nb)C wafers in 25.3% voided Square Lattice Honeycomb (SLHC). The H 2 propellant passes through these flow channels. This study is concerned with thermohydrodynamic analysis of the fuel element for propulsion with hypothetical axial power distribution because nuclear analysis of TERA has not been performed yet. As a result, when the power distribution of INSPI's M-SLHC is applied to the fuel assembly, the local heat concentration of fuel is more serious and the pressure of the initial inlet H 2 is higher than those of constant average power distribution applied. This means the fuel assembly geometry of 1.5 mm fuel wafers and 25.3% voided SLHC needs to be changed in order to reduce thermal and mechanical shocks. (author)

  11. 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.

  12. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    Ignatiev, V.; Devell, L.

    1995-01-01

    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

  13. FY 2017 – Thermal Aging Effects on Advanced Structural Materials

    Energy Technology Data Exchange (ETDEWEB)

    Li, Meimei [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K [Argonne National Lab. (ANL), Argonne, IL (United States); Chen, Wei-Ying [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-08-01

    This report provides an update on the evaluation of the effect of thermal aging on tensile properties of existing laboratory-sized heats of Alloy 709 austenitic stainless steel and the completion of effort on the thermal aging effect on the tensile properties of optimized G92 ferritic-martensitic steel. The report is a Level 3 deliverable in FY17 (M3AT-17AN1602081), under the Work Package AT-17AN160208, “Advanced Alloy Testing - ANL” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

  14. Proceedings of the 2004 international congress on advances in nuclear power plants - ICAPP'04

    International Nuclear Information System (INIS)

    2004-01-01

    Management; Ex-Vessel Debris Coolability and Steam Explosion: Theory and Modeling; Ex-Vessel Debris Coolability and Steam Explosion: Experiments and Supporting Analysis; PRA and Risk-informed Decision Making: Methodology; PRA and Risk-informed Decision Making: Advances in Practice; Use of CFD in Plant Safety Assessment and Related Regulatory Issues; Development and Application of Severe Accident Analysis Code); 6 - Thermal Hydraulic Analysis and Testing (Advances in Two-Phase Flow and Heat Transfer; Advances in CHF and Rod Bundle Thermal Hydraulics; CFD Applications to Water, Liquid Metal, and Gas Reactors; Separate Effects Thermal Hydraulic Experiments and Analysis; Integral Systems Thermal Hydraulic Experiments; Benchmark Analysis and Assessment; Natural Circulation Thermal Hydraulics; Thermal Striping and Thermal Stratification Studies); 7 - Core and Fuel Cycle Concepts and Experiments (Innovations in Core Designs; Advances in Core Design Methodology and Experimental Benchmarking; Advanced Fuel Cycles, Recycling, and Actinide Transmutation; Out of Core Fuel Cycle Issues); 8 - Material and Structural Issues (Structural and Materials Modeling and Analysis; Testing and Analysis of Structures and Materials; Advanced Issues in Welding and Materials; Fuel Design and Irradiation Issues for Next Generation Plants; Materials' Issues for Next Generation Plants); 9 - Nuclear Energy and Sustainability Including Hydrogen, Desalination, and Other Applications (Nuclear Energy Sustainability and Desalination; Nuclear Energy Application - Hydrogen); 10 - Space Power and Propulsion (Space Nuclear Power and Propulsion Systems; Nuclear Thermal Propulsion Concepts; Test and Design Methods; Instrumentation for Space Nuclear Reactors; Materials for Space Reactor Concepts)

  15. Turbopump options for nuclear thermal rockets

    International Nuclear Information System (INIS)

    Bissell, W.R.; Gunn, S.V.

    1992-07-01

    Several turbopump options for delivering liquid nitrogen to nuclear thermal rocket (NTR) engines were evaluated and compared. Axial and centrifugal flow pumps were optimized, with and without boost pumps, utilizing current design criteria within the latest turbopump technology limits. Two possible NTR design points were used, a modest pump pressure rise of 1,743 psia and a relatively higher pump pressure rise of 4,480 psia. Both engines utilized the expander cycle to maximize engine performance for the long duration mission. Pump suction performance was evaluated. Turbopumps with conventional cavitating inducers were compared with zero NPSH (saturated liquid in the tanks) pumps over a range of tank saturation pressures, with and without boost pumps. Results indicate that zero NSPH pumps at high tank vapor pressures, 60 psia, are very similar to those with the finite NPSHs. At low vapor pressures efficiencies fall and turbine pressure ratios increase leading to decreased engine chamber pressures and or increased pump pressure discharges and attendant high-pressure component weights. It may be concluded that zero tank NSPH capabilities can be obtained with little penalty to the engine systems but boost pumps are needed if tank vapor pressure drops below 30 psia. Axial pumps have slight advantages in weight and chamber pressure capability while centrifugal pumps have a greater operating range. 10 refs

  16. Overview of experimental work to ensure innovation of nuclear fuel for future advanced PWRs

    International Nuclear Information System (INIS)

    Zymak, J.; Valach, M.; Hejna, J.

    2002-11-01

    It is envisaged that advanced nuclear fuel will be operated in high burnup conditions, at a high linear power and at considerable mechanical fuel-cladding interactions. The report gives an overview of experimental work investigating phenomena that will affect APWR fuel, such as the manufacturing technology, thermal properties and safety requirements

  17. Helium mobility in advanced nuclear ceramics

    International Nuclear Information System (INIS)

    Agarwal, Shradha

    2014-01-01

    The main goal of this work is to improve our knowledge on the mechanisms able to drive the helium behaviour in transition metal carbides and nitrides submitted to thermal annealing or ion irradiation. TiC, TiN and ZrC polycrystals were implanted with 3 MeV 3 He ions at room temperature in the fluence range 2 * 10 15 et 6 * 10 16 cm -2 . Some of them have been pre-irradiated with self-ions (14 MeV Ti or Zr). Fully controlled thermal annealing tests were subsequently carried out in the temperature range 1000 - 1600 C for two hours. The evolution of the helium depth distribution in function of implantation dose, temperature and pre-irradiation dose was measured thanks to the deuteron-induced nuclear reaction 3 He(d, p 0 ) 4 He between 900 keV and 1.8 MeV. The microstructure of implanted and annealed samples was investigated by transmission electron microscopy on thin foils prepared using the FIB technique. Additional characterization tools, as X-ray diffraction and Raman microspectrometry, have been also applied in order to obtain complementary information. Among the most relevant results obtained, the following have to be outlined: - double-peak helium depth profile was measured on as implanted sample for the three compounds. The first peak is located near the end of range and includes the major part of helium, a second peak located close to the surface corresponds to the helium atoms trapped by the native vacancies; - the helium retention capacity in transition metal carbides and nitrides submitted to fully controlled thermal treatments varies according to ZrC 0.92 ≤ TiC 0.96 ≤ TiN 0.96 ; - whatever the investigated material, a self-ion-induced pre-damaging does not modify the initial helium profile extent. The influence of the post-implantation thermal treatment remains preponderant in any case; - the apparent diffusion coefficient of helium is in the range 4 * 10 -18 - 2 * 10 -17 m 2 s -1 in TiC0.96 and 3.5 * 10 -19 - 5.3 * 10 -18 m 2 s -1 in TiN 0.96 between

  18. Test facilities for evaluating nuclear thermal propulsion systems

    International Nuclear Information System (INIS)

    Beck, D.F.; Allen, G.C.; Shipers, L.R.; Dobranich, D.; Ottinger, C.A.; Harmon, C.D.; Fan, W.C.; Todosow, M.

    1992-01-01

    Interagency panels evaluating nuclear thermal propulsion (NTP) development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and baseline performance of some of the major subsystems designed to support a proposed ground test complex for evaluating nuclear thermal propulsion fuel elements and engines being developed for the Space Nuclear Thermal Propulsion (SNTP) program. Some preliminary results of evaluating this facility for use in testing other NTP concepts are also summarized

  19. Advances in nuclear fuel technology. 3. Development of advanced nuclear fuel recycle systems

    International Nuclear Information System (INIS)

    Arie, Kazuo; Abe, Tomoyuki; Arai, Yasuo

    2002-01-01

    Fast breeder reactor (FBR) cycle technology has a technical characteristics flexibly easy to apply to diverse fuel compositions such as plutonium, minor actinides, and so on and fuel configurations. By using this characteristics, various feasibilities on effective application of uranium resources based on breeding of uranium of plutonium for original mission of FBR, contribution to radioactive wastes problems based on amounts reduction of transuranium elements (TRU) in high level radioactive wastes, upgrading of nuclear diffusion resistance, extremely upgrading of economical efficiency, and so on. In this paper, were introduced from these viewpoints, on practice strategy survey study on FBR cycle performed by cooperation of the Japan Nuclear Cycle Development Institute (JNC) with electric business companies and so on, and on technical development on advanced nuclear fuel recycle systems carried out at the Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute, and so on. Here were explained under a vision on new type of fuels such as nitride fuels, metal fuels, and so on as well as oxide fuels, a new recycle system making possible to use actinides except uranium and plutonium, an 'advanced nuclear fuel cycle technology', containing improvement of conventional wet Purex method reprocessing technology, fuel manufacturing technology, and so on. (G.K.)

  20. Thermal Characterization of Nanostructures and Advanced Engineered Materials

    Science.gov (United States)

    Goyal, Vivek Kumar

    to heat-sinking units. This dissertation presents results of the experimental investigation and theoretical interpretation of thermal transport in the advanced engineered materials, which include thin films for thermal management of nanoscale devices, nanostructured superlattices as promising candidates for high-efficiency thermoelectric materials, and improved TIMs with graphene and metal particles as fillers providing enhanced thermal conductivity. The advanced engineered materials studied include chemical vapor deposition (CVD) grown ultrananocrystalline diamond (UNCD) and microcrystalline diamond (MCD) films on Si substrates, directly integrated nanocrystalline diamond (NCD) films on GaN, free-standing polycrystalline graphene (PCG) films, graphene oxide (GOx) films, and "pseudo-superlattices" of the mechanically exfoliated Bi2Te3 topological insulator films, and thermal interface materials (TIMs) with graphene fillers.

  1. Advanced nuclear reactor safety design technology research in NPIC

    International Nuclear Information System (INIS)

    Yu, H.

    2014-01-01

    After the Fukushima accident happen, Nuclear Power Plants (NPPs) construction has been suspended in China for a time. Now the new regulatory rule has been proposed that the most advanced safety standard must be adopted for the new NPPs and practical elimination of large fission product release by design during the next five plans period. So the advanced reactor research is developing in China. NPIC is engaging on the ACP1000 and ACP100 (Small Module Reactor) design. The main design character will be introduced in this paper. The Passive Combined with Active (PCWA) design was adopted during the ACP1000 design to reduce the core damage frequency (CDF); the Cavity Injection System (CIS) is design to mitigation the consequence of the severe accident. Advance passive safety system was designed to ensure the long term residual heat removal during the Small Module Reactor (SMR). The SMR will be utilized to be the floating reactors, district heating reactor and so on. Besides, the Science and Technology on Reactor System Design Technology Laboratory (LRSDT) also engaged on the fundamental thermal-hydraulic characteristic research in support of the system validation. (author)

  2. The thermal curve of nuclear matter

    International Nuclear Information System (INIS)

    Ma, Y.G.; Peter, J.; Siwek, A.; Bocage, F.; Bougault, R.; Brou, R.; Colin, J.; Cussol, D.; Durand, D.; Genouin-Duhamel, E.; Gulminelli, F.; Lecolley, J.F.; Lefort, T.; Le Neindre, N.; Lopez, O.; Louvel, M.; Nguyen, A.D.; Steckmeyer, J.C.; Tamain, B.; Vient, E.

    1997-01-01

    Earlier measurements of nuclear matter thermal curve of liquid to gas phase transition presented two limitation: only one temperature measuring method was available and the mass number of the formed nuclei decreased from 190 to 50 when the excitation energy increased. To avoid these limitations experiments with the multidetector INDRA at GANIL were carried-out. Among the quasi-projectiles issued from the 36 Ar collisions at 52, 74, 95 A.MeV on the 58 Ni, nuclei of close masses were selected. The excitation energy was determined by the calorimetry of the charged products emitted by quasi-projectiles while the temperature was measured by three different methods. Very different apparent temperatures were obtained for the same excitation energy/nucleon. Only one curve displays a slope variation but no indication of plateau. With the quasi-projectiles obtained from the collisions of 129 Xe at 50 MeV/u on a 119 Sn target behaviors similar to those of 36 Ar were observed in the covered domain of excitation energy. To solve this puzzle and recover the initial temperatures of interest the only mean was to do a theoretical simulation in which one follows the de-excitation of the nuclei formed at different excitation energies and look for the thermal curve able to reproduce the observed temperatures. Two extreme possibilities were taken into account concerning the de-excitation process: either a sequential process established at E * /A≤ 3 MeV/u or a sudden multifragmentation in several hot fragments, most probably at E * /A≥ 10 MeV/u. In both cases it was possible to reproduce the whole set of experimental results concerning the 36 Ar projectile. The initial temperature increases steadily as a function of excitation energy showing no plateau or singular points. The results indicate that, being a system without external pressure, in its passage from the liquid phase to the gas phase the nucleus does not display necessarily a temperature plateau. Discussions on

  3. Superconducting Electric Boost Pump for Nuclear Thermal Propulsion, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — A submersible, superconducting electric boost pump sized to meet the needs of future Nuclear Thermal Propulsion systems in the 25,000 lbf thrust range is proposed....

  4. Proceedings of national symposium on advanced instrumentation for nuclear research

    International Nuclear Information System (INIS)

    1993-01-01

    The National Symposium on Advanced Instrumentation for Nuclear Research was held in Bombay during January 27-29, 1993 at BARC. Progress of modern nuclear research is closely related to the availability of state of the art instruments and systems. With the advancements in experimental techniques and sophisticated detector developments, the performance specifications have become more stringent. State of the art techniques and diverse applications of sophisticated nuclear instrumentation systems are discussed along with indigenous efforts to meet the specific instrumentation needs of research programs in nuclear sciences. Papers of relevance to nuclear science and technology are indexed separately. (original)

  5. Advanced fuels for nuclear fusion reactors

    International Nuclear Information System (INIS)

    McNally, J.R. Jr.

    1974-01-01

    Should magnetic confinement of hot plasma prove satisfactory at high β (16 πnkT//sub B 2 / greater than 0.1), thermonuclear fusion fuels other than D.T may be contemplated for future fusion reactors. The prospect of the advanced fusion fuels D.D and 6 Li.D for fusion reactors is quite promising provided the system is large, well reflected and possesses a high β. The first generation reactions produce the very active, energy-rich fuels t and 3 He which exhibit a high burnup probability in very hot plasmas. Steady state burning of D.D can ensue in a 60 kG field, 5 m reactor for β approximately 0.2 and reflectivity R/sub mu/ = 0.9 provided the confinement time is about 38 sec. The feasibility of steady state burning of 6 Li.D has not yet been demonstrated but many important features of such systems still need to be incorporated in the reactivity code. In particular, there is a need for new and improved nuclear cross section data for over 80 reaction possibilities

  6. TALSPEAK Chemistry in Advanced Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    Nilsson, Mikael; Nash, Kenneth L.

    2008-01-01

    The separation of trivalent transplutonium actinides from fission product lanthanide ions represents a challenging aspect of advanced nuclear fuel partitioning schemes. The challenge of this separation could be amplified in the context of the AFCI-UREX+1a process, as Np and Pu will accompany the minor actinides to this stage of separation. At present, the baseline lanthanide-actinide separation method is the TALSPEAK (Trivalent Actinide - Lanthanide Separation by Phosphorus reagent Extraction from Aqueous complexes) process. TALSPEAK was developed in the late 1960's at Oak Ridge National Laboratory and has been demonstrated at pilot scale. This process relies on the complex interaction between an organic and an aqueous phase both containing complexants for selectively separating the trivalent actinide. The 3 complexing components are: the di(2-ethylhexyl) phosphoric acid (HDEHP), the lactic acid (HL) and the diethylenetriamine-N,N,N',N'',N''-pentaacetic acid (DTPA). In this report we discuss observations on kinetic and thermodynamic features described in the prior literature and describe some results of our ongoing research on basic chemical features of this system. The information presented indicates that the lactic acid buffer participates in the net operation of the TALSPEAK process in a manner that is not explained by existing information on the thermodynamic features if the known Eu(III)-lactate species. (authors)

  7. Proceedings of the 10th international topical meeting on nuclear thermal hydraulics, operation and safety (NUTHOS-10)

    International Nuclear Information System (INIS)

    2014-01-01

    The 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-10) in Okinawa, Japan is sponsored by Atomic Energy Society of Japan, in cooperation with the International Atomic Energy Agency, and co-sponsored by American Nuclear Society Thermal Hydraulics Division among others. Enhanced safety and reducing cost are going together, which can be achieved through continued research and development efforts. NUTHOS keeps you abreast of the most updated information in the advancement of science and technology in nuclear thermal hydraulics, operations and safety, and provides you insights into the future. (J.P.N.)

  8. Thermal Model Predictions of Advanced Stirling Radioisotope Generator Performance

    Science.gov (United States)

    Wang, Xiao-Yen J.; Fabanich, William Anthony; Schmitz, Paul C.

    2014-01-01

    This paper presents recent thermal model results of the Advanced Stirling Radioisotope Generator (ASRG). The three-dimensional (3D) ASRG thermal power model was built using the Thermal Desktop(trademark) thermal analyzer. The model was correlated with ASRG engineering unit test data and ASRG flight unit predictions from Lockheed Martin's (LM's) I-deas(trademark) TMG thermal model. The auxiliary cooling system (ACS) of the ASRG is also included in the ASRG thermal model. The ACS is designed to remove waste heat from the ASRG so that it can be used to heat spacecraft components. The performance of the ACS is reported under nominal conditions and during a Venus flyby scenario. The results for the nominal case are validated with data from Lockheed Martin. Transient thermal analysis results of ASRG for a Venus flyby with a representative trajectory are also presented. In addition, model results of an ASRG mounted on a Cassini-like spacecraft with a sunshade are presented to show a way to mitigate the high temperatures of a Venus flyby. It was predicted that the sunshade can lower the temperature of the ASRG alternator by 20 C for the representative Venus flyby trajectory. The 3D model also was modified to predict generator performance after a single Advanced Stirling Convertor failure. The geometry of the Microtherm HT insulation block on the outboard side was modified to match deformation and shrinkage observed during testing of a prototypic ASRG test fixture by LM. Test conditions and test data were used to correlate the model by adjusting the thermal conductivity of the deformed insulation to match the post-heat-dump steady state temperatures. Results for these conditions showed that the performance of the still-functioning inboard ACS was unaffected.

  9. Influence of prolonged nuclear fuel burnup on safety characteristics of advanced PWRs

    International Nuclear Information System (INIS)

    Spasojevic, D.; Matausek, M.; Marinkovic, N.

    1989-01-01

    Prolonged nuclear fuel burnup in advanced NPP with four or more instead of three one-year cycles, and/or with 15- to 18-month instead of standard 12-month cycles, requires the fresh fuel to have increased enrichment combined with burnable poisons. This causes changes in axial and radial distribution of power generation during the particular fuel cycles, so that detailed analyses of thermal reliability of reactor core becomes necessary. This paper presents the results of the analysis of the departure from nuclear boiling ratio DNBR for an equilibrium cycle of an advanced PWR. (author)

  10. Thermal analysis of transportation packaging for nuclear spent fuel

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki

    1989-01-01

    Safety analysis of transportation packaging for nuclear spent fuel comprises structural, thermal, containment, shielding and criticality factors, and the safety of a packaging is verified by these analyses. In thermal analysis, the temperature of each part of the packaging is calculated under normal and accident test conditions. As an example of thermal analysis, the temperature distribution of a packaging being subjected to a normal test was calculated by the TRUMP code and compared with measured data. (author)

  11. On the thermal properties of polarized nuclear matter

    International Nuclear Information System (INIS)

    Hassan, M.Y.M.; Montasser, S.S.; Ramadan, S.

    1979-08-01

    The thermal properties of polarized nuclear matter are calculated using Skyrme III interaction modified by Dabrowski for polarized nuclear matter. The temperature dependence of the volume, isospin, spin and spin isospin pressure and energies are determined. The temperature, isospin, spin and spin isospin dependence of the equilibrium Fermi momentum is also discussed. (author)

  12. On the spin saturation and thermal properties of nuclear matter

    International Nuclear Information System (INIS)

    Hassan, M.Y.M.; Ramadan, S.

    1983-12-01

    The binding energy and the incompressibility of nuclear matter with degree of spin saturation D is calculated using the Skyrme interaction and two forms of a velocity dependent effective potential. The effect of the degree of spin saturation D on the thermal properties of nuclear matter is also discussed. It is found that generally the pressure decreases with increasing D. (author)

  13. Nuclear science and metallurgy. Advances and interactions

    International Nuclear Information System (INIS)

    Grison, Emmanuel

    1977-01-01

    The history of the production of atomic power by fission since 1953 is reviewed: metallurgy of uranium; the new metals zirconium and beryllium; steels and nuclear structure; nuclear reactor vessels; water corrosion; effects of radiations [fr

  14. Nuclear science and metallurgy. Advances and interactions

    Energy Technology Data Exchange (ETDEWEB)

    Grison, E [CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1977-01-01

    The history of the production of atomic power by fission since 1953 is reviewed: metallurgy of uranium; the new metals zirconium and beryllium; steels and nuclear structure; nuclear reactor vessels; water corrosion; effects of radiations.

  15. Construction and engineering report for advanced nuclear fuel development facility

    International Nuclear Information System (INIS)

    Cho, S. W.; Park, J. S.; Kwon, S.J.; Lee, K. W.; Kim, I. J.; Yu, C. H.

    2003-09-01

    The design and construction of the fuel technology development facility was aimed to accommodate general nuclear fuel research and development for the HANARO fuel fabrication and advanced fuel researches. 1. Building size and room function 1) Building total area : approx. 3,618m 2 , basement 1st floor, ground 3th floor 2) Room function : basement floor(machine room, electrical room, radioactive waste tank room), 1st floor(research reactor fuel fabrication facility, pyroprocess lab., metal fuel lab., nondestructive lab., pellet processing lab., access control room, sintering lab., etc), 2nd floor(thermal properties measurement lab., pellet characterization lab., powder analysis lab., microstructure analysis lab., etc), 3rd floor(AHU and ACU Room) 2. Special facility equipment 1) Environmental pollution protection equipment : ACU(2sets), 2) Emergency operating system : diesel generator(1set), 3) Nuclear material handle, storage and transport system : overhead crane(3sets), monorail hoist(1set), jib crane(2sets), tank(1set) 4) Air conditioning unit facility : AHU(3sets), packaged air conditioning unit(5sets), 5) Automatic control system and fire protection system : central control equipment(1set), lon device(1set), fire hose cabinet(3sets), fire pump(3sets) etc

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

    International Nuclear Information System (INIS)

    Alameri, Saeed A.; King, Jeffrey C.

    2013-01-01

    Nuclear power plants operate most economically at a constant power level, providing base load electric power. In an energy grid containing a high fraction of renewable power sources, nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling a nuclear reactor to a large thermal energy storage block will allow the reactor to better respond to variable power demands. In the system described in this paper, a Prismatic core Advanced High Temperature Reactor supplies constant power to a lithium chloride molten salt thermal energy storage block that provides thermal power as needed to a closed Brayton cycle energy conversion system. During normal operation, the thermal energy storage block stores thermal energy during the night for use in the times of peak demand during the day. In this case, the nuclear reactor stays at a constant thermal power level. After a loss of forced circulation, the reactor reaches a shut down state in less than half an hour and the average fuel, graphite and coolant temperatures remain well within the design limits over the duration of the transient, demonstrating the inherent safety of the coupled system. (author)

  17. Application of thermal-hydraulic codes in the nuclear sector

    International Nuclear Information System (INIS)

    Queral, C.; Coriso, M.; Garcia Sedano, P. J.; Ruiz, J. A.; Posada, J. M.; Jimenez Varas, G.; Sol, I.; Herranz, L. E.

    2011-01-01

    Use of thermal-hydraulic codes is extended all over many different aspects of nuclear engineering. This article groups and briefly describes the main features of some of the well known codes as an introduction to their recent applications in the Spain nuclear sector. the broad range and quality of applications highlight the maturity achieved both in industry and research organizations and universities within the Spanish nuclear sector. (Author)

  18. Preliminary development of thermal nuclear cell homogenization code

    International Nuclear Information System (INIS)

    Su'ud, Z.; Shafii, M. A.; Yudha, S. P.; Waris, A.; Rijal, K.

    2012-01-01

    Nuclear fuel cell homogenization for thermal reactors usually include three main parts, i.e., fast energy resonance part which usually adopt narrow resonance approximation to treat the resonance, low (intermediate) energy region in which the resonance can not be treated accurately using NR approximation and therefore we should use intermediate resonance treatment, and thermal energy region (very low) in which the effect of thermal must be treated properly. In n this study the application of the intermediate resonance approximation treatment for low energy nuclear resonance is discussed. The method is iterative based. As a sample the method is applied in U-235 low lying resonance and the result is presented and discussed.

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

    International Nuclear Information System (INIS)

    McEligot, D.M.; Condie, K.G.; Foust, T.D.; McCreery, G.E.; Pink, R.J.; Stacey, D.E.; Shenoy, A.; Baccaglini, G.; Pletcher, R.H.; Wallace, J.M.; Vukoslavcevic, P.; Jackson, J.D.; Kunugi, T.; Satake, S.-I.

    2002-01-01

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

  20. Design related aspects in advanced nuclear fission plants

    International Nuclear Information System (INIS)

    Hoffelner, Wolfgang

    2011-01-01

    Important issues to be considered for design of future reactors are: extrapolation of stress rupture data, creep-fatigue, negligible creep, damage monitoring. The paper highlights some new developments taking examples from a martensitic steel (mod 9% Cr), oxide dispersion strengthened (ODS) steels and nickel-base superalloys. Traditional approaches to extrapolation of (thermal) stress rupture data like Larson-Miller Parameter or Monkman-Grant rule seem to be valid concepts also for advanced reactors. However, a significant influence of cyclic softening on creep rates and stress rupture data can be expected as shown for grade 91. This is particularly true for creep-fatigue interactions. Based on cyclic stress-strain behaviour it is also possible to get very good life-time predictions under creep-fatigue with a strain range separation (inelastic fatigue and creep ranges) technique which could replace the currently used linear life fraction rule. Results from in-beam irradiation creep reveal no significant influence of dispersoid size. It can be assumed that irradiation creep is a matrix property. Finally it is shown that micro-sample testing of exposed material could be used as an advanced method for damage assessment in future nuclear power plants.

  1. Non-proliferation and advances in nuclear science

    International Nuclear Information System (INIS)

    Iyengar, P.K.

    1995-01-01

    So far, the non-proliferation treaty (NPT) has concentrated on safeguard regimes based on technologies relating to the production of uranium and plutonium in nuclear reactors, and on their potential diversion for use in nuclear weapons. As nuclear science advances, however, nuclear technology both peaceful and for weapons will change, and for the NPT to remain relevant, it must reflect these changes. At this juncture, when the NPT is coming up for review in a year's time, it is important for physicists to take a fresh look at recent advances in nuclear science, and inform the policy-makers and the public at large about their potential for impacting nuclear technology in the future. In this article a few such advances are highlighted and their implications for the NPT are considered. (author). 4 refs

  2. Effective thermal conductivity of advanced ceramic breeder pebble beds

    Energy Technology Data Exchange (ETDEWEB)

    Pupeschi, S., E-mail: simone.pupeschi@kit.edu; Knitter, R.; Kamlah, M.

    2017-03-15

    As the knowledge of the effective thermal conductivity of ceramic breeder pebble beds under fusion relevant conditions is essential for the development of solid breeder blanket concepts, the EU advanced and reference lithium orthosilicate material were investigated with a newly developed experimental setup based on the transient hot wire method. The effective thermal conductivity was investigated in the temperature range RT–700 °C. Experiments were performed in helium and air atmospheres in the pressure range 0.12–0.4 MPa (abs.) under a compressive load up to 6 MPa. Results show a negligible influence of the chemical composition of the solid material on the bed’s effective thermal conductivity. A severe reduction of the effective thermal conductivity was observed in air. In both atmospheres an increase of the effective thermal conductivity with the temperature was detected, while the influence of the compressive load was found to be small. A clear dependence of the effective thermal conductivity on the pressure of the filling gas was observed in helium in contrast to air, where the pressure dependence was drastically reduced.

  3. Emission and thermal performance upgrade through advanced control backfit

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, A.K. [Stone & Webster Engineering Corporation, Boston, MA (United States)

    1994-12-31

    Reducing emission and improving thermal performance of currently operating power plants is a high priority. A majority of these power plants are over 20 years old with old control systems. Upgrading the existing control systems with the latest technology has many benefits, the most cost beneficial are the reduction of emission and improving thermal performance. The payback period is usually less than two years. Virginia Power is installing Stone & Webster`s NO{sub x} Emissions Advisor and Advanced Steam Temperature Control systems on Possum Point Units 3 and 4 to achieve near term NO{sub x} reductions while maintaining high thermal performance. Testing has demonstrated NO{sub x} reductions of greater than 20 percent through the application of NO{sub x} Emissions Advisor on these units. The Advanced Steam Temperature Control system which has been operational at Virginia Power`s Mt. Storm Unit 1 has demonstrated a signification improvement in unit thermal performance and controllability. These control systems are being combined at Units 3 and 4 to reduce NO{sub x} emissions and achieve improved unit thermal performance and control response with the existing combustion hardware. Installation has been initiated and is expected to be completed by the spring of 1995. Possum Point Power Station Units 3 and 4 are pulverized coal, tangentially fired boilers producing 107 and 232 MW and have a distributed control system and a PC based performance monitoring system. The installation of the advanced control and automation system will utilize existing control equipment requiring the addition of several PCs and PLC.

  4. Experiences and Trends of Manufacturing Technology of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    2012-08-01

    The 'Atoms for Peace' mission initiated in the mid-1950s paved the way for the development and deployment of nuclear fission reactors as a source of heat energy for electricity generation in nuclear power reactors and as a source of neutrons in non-power reactors for research, materials irradiation, and testing and production of radioisotopes. The fuels for nuclear reactors are manufactured from natural uranium (∼99.3% 238 U + ∼0.7% 235 U) and natural thorium (∼100% 232 Th) resources. Currently, most power and research reactors use 235 U, the only fissile isotope found in nature, as fuel. The fertile isotopes 238 U and 232 Th are transmuted in the reactor to human-made 239 Pu and 233 U fissile isotopes, respectively. Likewise, minor actinides (MA) (Np, Am and Cm) and other plutonium isotopes are also formed by a series of neutron capture reactions with 238 U and 235 U. Long term sustainability of nuclear power will depend to a great extent on the efficient, safe and secure utilization of fissile and fertile materials. Light water reactors (LWRs) account for more than 82% of the operating reactors, followed by pressurized heavy water reactors (PHWRs), which constitute ∼10% of reactors. LWRs will continue to dominate the nuclear power market for several decades, as long as economically viable natural uranium resources are available. Currently, the plutonium obtained from spent nuclear fuel is subjected to mono recycling in LWRs as uranium-plutonium mixed oxide (MOX), containing up to 12% PuO 2 , in a very limited way. The reprocessed uranium (RepU) is also re-enriched and recycled in LWRs in a few countries. Unfortunately, the utilization of natural uranium resources in thermal neutron reactors is 2 and MOX fuel technology has matured during the past five decades. These fuels are now being manufactured, used and reprocessed on an industrial scale. Mixed uranium- plutonium monocarbide (MC), mononitride (MN) and U-Pu-Zr alloys are recognized as advanced fuels

  5. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    Science.gov (United States)

    Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

    2012-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

  6. Advanced nuclear reactors and their simulators

    International Nuclear Information System (INIS)

    Chaushevski, Anton; Boshevski, Tome

    2003-01-01

    Population growth, economy development and improvement life standard impact on continually energy needs as well as electricity. Fossil fuels have limited reserves, instability market prices and destroying environmental impacts. The hydro energy capacities highly depend on geographic and climate conditions. The nuclear fission is significant factor for covering electricity needs in this century. Reasonable capital costs, low fuel and operating expenses, environmental acceptable are some of the facts that makes the nuclear energy an attractive option especially for the developing countries. The simulators for nuclear reactors are an additional software tool in order to understand, study research and analyze the processes in nuclear reactors. (Original)

  7. Advance of nuclear cardiology in clinical application

    International Nuclear Information System (INIS)

    Shi Hongcheng

    2007-01-01

    Nuclear cardiology has make a little bit progress in the past year. Both nuclear cardiology and other cardiac imaging have its own advantage and disadvantage in the diagnostic of coronary artery disease. And the relationship of them is complementary but not instead of each other. Nuclear cardiology provides a one-stop shop for diagnosis, risk stratification, and management of coronary artery disease. Nuclear cardiology plays a very important role in the diagnostic of coronary artery disease in early stage in the special group of people. (authors)

  8. Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

    2010-01-01

    The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

  9. Workshop on advanced nuclear data online services. Summary report

    International Nuclear Information System (INIS)

    Schwerer, O.

    2000-09-01

    This report summarizes the Workshop on Advanced Nuclear Data Online Services held in Vienna, Austria, 29 November to 3 December 1999. It includes the main topics of the agenda, the list of participants, and general comments and recommendations. (author)

  10. Thermal coupling system analysis of a nuclear desalination plant

    International Nuclear Information System (INIS)

    Adak, A.K.; Srivastava, V.K.; Tewari, P.K.

    2010-01-01

    When a nuclear reactor is used to supply steam for desalination plant, the method of coupling has a significant technical and economic impact. The exact method of coupling depends upon the type of reactor and type of desalination plant. As a part of Nuclear Desalination Demonstration Project (NDDP), BARC has successfully commissioned a 4500 m 3 /day MSF desalination plant coupled to Madras Atomic Power Station (MAPS) at Kalpakkam. Desalination plant coupled to nuclear power plant of Pressurized Heavy Water Reactor (PHWR) type is a good example of dual-purpose nuclear desalination plant. This paper presents the thermal coupling system analysis of this plant along with technical and safety aspects. (author)

  11. Advanced nuclear control and protection system ANCAP-80

    International Nuclear Information System (INIS)

    Asai, Takashi; Okano, Michihiko; Ishibashi, Kengo; Hasegawa, Masakoto; Fukuda, Hiroyoshi; Hosomichi, Renichi.

    1983-01-01

    Advanced reactor protection systems were developed to improve operational reliability and availability and to ease the burden of operators of Mitsubishi PWR Nuclear Power Stations. (Called ANCAP-80; Advanced Nuclear Control And Protection System) For the PWR plants now being planned and in future plans, Mitsubishi will adopt these systems with the following functional features; (1) Four channel protection logic, (2) Automatic bypass logic, (3) Automatic test provision, (4) Optical isolators. (author)

  12. solar thermal power systems advanced solar thermal technology project, advanced subsystems development

    Science.gov (United States)

    1979-01-01

    The preliminary design for a prototype small (20 kWe) solar thermal electric generating unit was completed, consisting of several subsystems. The concentrator and the receiver collect solar energy and a thermal buffer storage with a transport system is used to provide a partially smoothed heat input to the Stirling engine. A fossil-fuel combustor is included in the receiver designs to permit operation with partial or no solar insolation (hybrid). The engine converts the heat input into mechanical action that powers a generator. To obtain electric power on a large scale, multiple solar modules will be required to operate in parallel. The small solar electric power plant used as a baseline design will provide electricity at remote sites and small communities.

  13. Advances in instrumentation for nuclear astrophysics

    Directory of Open Access Journals (Sweden)

    S. D. Pain

    2014-04-01

    Full Text Available The study of the nuclear physics properties which govern energy generation and nucleosynthesis in the astrophysical phenomena we observe in the universe is crucial to understanding how these objects behave and how the chemical history of the universe evolved to its present state. The low cross sections and short nuclear lifetimes involved in many of these reactions make their experimental determination challenging, requiring developments in beams and instrumentation. A selection of developments in nuclear astrophysics instrumentation is discussed, using as examples projects involving the nuclear astrophysics group at Oak Ridge National Laboratory. These developments will be key to the instrumentation necessary to fully exploit nuclear astrophysics opportunities at the Facility for Rare Isotope Beams which is currently under construction.

  14. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Prodea, Iosif; Catana, Alexandru

    2010-01-01

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  15. Performance testing of thermal analysis codes for nuclear fuel casks

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

  16. The global greenhouse effect and the advanced nuclear energy system

    International Nuclear Information System (INIS)

    Byong Whi Lee

    1998-01-01

    In spite of future uncertainty, Korea is very much committed to nuclear energy as a major source of electric power expansion, because of its lack of domestic energy resources. A long term nuclear power program has resulted in 11 nuclear power plants of 9.6 GWe in operation, 2 units under construction and 7 planned. This means that the share of nuclear power in Korean electricity production would be about 38% in 2006. Many other countries were faced with the problem of global warming which is related to carbondioxide emission from the use of fossil fuels. According to Korean experience, it could be concluded that substitution of fossil fuels would be the most efficient and economic means of reducing the greenhouse gas emissions. In addition to nuclear and hydropower, the most promising other non-fossil sources are geothermal energy, biomass, solar thermal energy, photovoltaic systems, wind power, tidal power, wave power and ocean thermal electric conversion

  17. Holistic safety analysis for advanced nuclear power plants

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Guimaraes, A.C.F.

    1992-01-01

    This paper reviews the basic methodology of safety analysis used in the ANGRA-I and ANGRA-II nuclear power plants, its weaknesses, the problems with public acceptance of the risks, the future of the nuclear energy in Brazil, as well as recommends a new methodology, HOLISTIC SAFETY ANALYSIS, to be used both in the design and licensing phases, for advanced reactors. (author)

  18. Advanced nuclear reactor safety issues and research needs

    International Nuclear Information System (INIS)

    2002-01-01

    On 18-20 February 2002, the OECD Nuclear Energy Agency (NEA) organised, with the co-sponsorship of the International Atomic Energy Agency (IAEA) and in collaboration with the European Commission (EC), a Workshop on Advanced Nuclear Reactor Safety Issues and Research Needs. Currently, advanced nuclear reactor projects range from the development of evolutionary and advanced light water reactor (LWR) designs to initial work to develop even further advanced designs which go beyond LWR technology (e.g. high-temperature gas-cooled reactors and liquid metal-cooled reactors). These advanced designs include a greater use of advanced technology and safety features than those employed in currently operating plants or approved designs. The objectives of the workshop were to: - facilitate early identification and resolution of safety issues by developing a consensus among participating countries on the identification of safety issues, the scope of research needed to address these issues and a potential approach to their resolution; - promote the preservation of knowledge and expertise on advanced reactor technology; - provide input to the Generation IV International Forum Technology Road-map. In addition, the workshop tried to link advancement of knowledge and understanding of advanced designs to the regulatory process, with emphasis on building public confidence. It also helped to document current views on advanced reactor safety and technology, thereby contributing to preserving knowledge and expertise before it is lost. (author)

  19. Addressing Attribution - Advances in Nuclear Forensics

    International Nuclear Information System (INIS)

    Luetzenkirchen, K.; Mayer, K.; Wallenius, M.; Varga, Z.

    2013-01-01

    In nuclear forensic investigations the information is typically used to find out the origin of the material including its production place and process. As nuclear material is an industrial product, the respective production process will unavoidably leave a 'fingerprint' in the material. In consequence, parameters related to process and source material are measured and conclusions about the history and origin of the nuclear material can be drawn. Chemical impurities, microstructure, molecular structure or isotopic composition are examples of such parameters, which compose a characteristic 'signature'. The rare-earth elemental (REE) pattern has been found to be an important signature to determine the type of ore used as source material for the production of uranium ore concentrate, since the REE pattern is invariable in most hydrometallurgical processes. Generally, Decrypting the information for nuclear forensic investigations requires sophisticated methodology, subject matter expertise and reference data. The paper is followed by the slides of the presentation. (A.C.)

  20. Nuclear fuel: modelling the advanced plutonium assembly

    International Nuclear Information System (INIS)

    Kaoua, Th.; Lenain, R.

    2004-01-01

    The benefits of modeling in the nuclear sector are illustrated by the example of the design study for a new plutonium fuel assembly, APA, capable of ensuring maximum consumption of this fuel in pressurized-water reactors. Beyond the physical design of the assembly and its integration into the reactor, this serves for the working out of a complete materials flow and assists in modeling production from the entire inventory of nuclear power stations. (authors)

  1. Nuclear fuel: modelling the advanced plutonium assembly

    International Nuclear Information System (INIS)

    N'kaoua, Th.; Lenain, R.

    2002-01-01

    The benefits of modeling in the nuclear sector are illustrated by the example of the design study for a new plutonium fuel assembly, APA, capable of ensuring maximum consumption of this fuel in pressurized-water reactors. Beyond the physical design of the assembly and its integration into the reactor, this serves for the working out of a complete materials flow and assists in modeling production from the entire inventory of nuclear power stations. (authors)

  2. Handling effluent from nuclear thermal propulsion system ground tests

    International Nuclear Information System (INIS)

    Shipers, L.R.; Allen, G.C.

    1992-01-01

    A variety of approaches for handling effluent from nuclear thermal propulsion system ground tests in an environmentally acceptable manner are discussed. The functional requirements of effluent treatment are defined and concept options are presented within the framework of these requirements. System concepts differ primarily in the choice of fission-product retention and waste handling concepts. The concept options considered range from closed cycle (venting the exhaust to a closed volume or recirculating the hydrogen in a closed loop) to open cycle (real time processing and venting of the effluent). This paper reviews the different methods to handle effluent from nuclear thermal propulsion system ground tests

  3. Effluent treatment options for nuclear thermal propulsion system ground tests

    International Nuclear Information System (INIS)

    Shipers, L.R.; Brockmann, J.E.

    1992-01-01

    A variety of approaches for handling effluent from nuclear thermal propulsion system ground tests in an environmentally acceptable manner are discussed. The functional requirements of effluent treatment are defined and concept options are presented within the framework of these requirements. System concepts differ primarily in the choice of fission-product retention and waste handling concepts. The concept options considered range from closed cycle (venting the exhaust to a closed volume or recirculating the hydrogen in a closed loop) to open cycle (real time processing and venting of the effluent). This paper reviews the strengths and weaknesses of different methods to handle effluent from nuclear thermal propulsion system ground tests

  4. Engine cycle design considerations for nuclear thermal propulsion systems

    International Nuclear Information System (INIS)

    Pelaccio, D.G.; Scheil, C.M.; Collins, J.T.

    1993-01-01

    A top-level study was performed which addresses nuclear thermal propulsion system engine cycle options and their applicability to support future Space Exploration Initiative manned lunar and Mars missions. Technical and development issues associated with expander, gas generator, and bleed cycle near-term, solid core nuclear thermal propulsion engines are identified and examined. In addition to performance and weight the influence of the engine cycle type on key design selection parameters such as design complexity, reliability, development time, and cost are discussed. Representative engine designs are presented and compared. Their applicability and performance impact on typical near-term lunar and Mars missions are shown

  5. Spanish program of advanced Nuclear Power Plants

    International Nuclear Information System (INIS)

    Marco, M.; Redon, R.

    1993-01-01

    The energy Spanish Plan is promoting some actions within the area of advanced reactors. Efforts are focussed onto the European Program of Advanced Reactors, the Program of Passive Plants (EPRI), European Fast Reactor Project and the APWR-1000 Program of INI. Electrical sector utilities and industrial partners supported by the Administration have organized an steering committee. The program of Passive Plants includes activities on Qualification, design and detailed engineering (Qualification project, SBWR project of G.E. and AP600 Project of Westinghouse. The european project on advanced plants has the following Spanish contribution: Analysis of alternative Dossier on European requisites (EUR) and Design of an European Reactor (EPR)

  6. Advances in chemical engineering in nuclear and process industries

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-06-01

    Symposium on Advances in Chemical Engineering in Nuclear and Process Industries dealt with a wide spectrum of areas encompassing various industries such as nuclear, fertilizer, petrochemical, refinery and cement. The topics covered in the symposium dealt with the advancements in the existing fields of science and technologies as well as in some of the emerging technologies such as membrane technology, bio-chemical and photo-chemical engineering etc. with a special emphasis on nuclear related aspects. Papers relevant to INIS are indexed separately.

  7. Advances in chemical engineering in nuclear and process industries

    International Nuclear Information System (INIS)

    1994-06-01

    Symposium on Advances in Chemical Engineering in Nuclear and Process Industries dealt with a wide spectrum of areas encompassing various industries such as nuclear, fertilizer, petrochemical, refinery and cement. The topics covered in the symposium dealt with the advancements in the existing fields of science and technologies as well as in some of the emerging technologies such as membrane technology, bio-chemical and photo-chemical engineering etc. with a special emphasis on nuclear related aspects. Papers relevant to INIS are indexed separately

  8. Coupled fast-thermal system at the 'RB' nuclear reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    1987-04-01

    The results of the analyses of the possibility of the coupled fast-thermal system (CFTS) design at the 'RB' nuclear reactor are shown. As the proof of the theoretical analyses the first stage CFTS-1 has been designed, realized, and tested. The excellent agreement between the results of the CFTS-1 studies and the theoretical predictions opens a straight way to the second, the final stage - realization of the designed CFST at the 'RB' nuclear reactor. (author)

  9. Technology readiness levels for advanced nuclear fuels and materials development

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W.J., E-mail: jon.carmack@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Braase, L.A.; Wigeland, R.A. [Idaho National Laboratory, Idaho Falls, ID (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States)

    2017-03-15

    Highlights: • Definition of nuclear fuels system technology readiness level. • Identification of evaluation criteria for nuclear fuel system TRLs. • Application of TRLs to fuel systems. - Abstract: The Technology Readiness process quantitatively assesses the maturity of a given technology. The National Aeronautics and Space Administration (NASA) pioneered the process in the 1980s to inform the development and deployment of new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications. It was also adopted by the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is needed to improve the performance and safety of current and advanced reactors, and ultimately close the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the assessment process to advanced fuel development is useful as a management, communication, and tracking tool. This article provides definition of technology readiness levels (TRLs) for nuclear fuel technology as well as selected examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).

  10. Perspective of nuclear energy and advanced reactors

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Cobian, J.

    2007-01-01

    Future nuclear energy growth will be the result of the contributions of every single plant being constructed or projected at present as it is connected to the grid. As per IAEA, there exists presently 34 nuclear power plants under construction 81 with the necessary permits and funding and 223 proposed, which are plants seriously pursuing permits and financing. This means that in a few decades the number of nuclear power plants in operation will have doubled. This growth rate is characterised by the incorporation of new countries to the nuclear club and the gradually increasing importance of Asian countries. During this expansive phase, generation III and III+designs are or will be used. These designs incorporate the experience from operating plants, and introduce innovations on rationalization design efficiency and safety, with emphasis on passive safety features. In a posterior phase, generation IV designs, presently under development, will be employed. Generation IV consists of several types of reactors (fast reactors, very high temperature reactors, etc), which will improve further sustain ability, economy, safety and reliability concepts. The described situation seems to lead to a renaissance of the nuclear energy to levels hardly thinkable several years ago. (Author)

  11. Equipping simulators with an advanced thermal hydraulics model EDF's experience

    International Nuclear Information System (INIS)

    Soldermann, R.; Poizat, F.; Sekri, A.; Faydide, B.; Dumas, J.M.

    1997-01-01

    The development of an accelerated version of the advanced CATHARe-1 thermal hydraulics code designed for EDF training simulators (CATHARE-SIMU) was successfully completed as early as 1991. Its successful integration as the principal model of the SIPA Post-Accident Simulator meant that its use could be extended to full-scale simulators as part of the renovation of the stock of existing simulators. In order to further extend the field of application to accidents occurring in shutdown states requiring action and to catch up with developments in respect of the CATHARE code, EDF initiated the SCAR Project designed to adapt CATHARE-2 to simulator requirements (acceleration, parallelization of the computation and extension of the simulation range). In other respects, the installation of SIPA on workstations means that the authors can envisage the application of this remarkable training facility to the understanding of thermal hydraulics accident phenomena

  12. Thermal hydraulic evaluation of advanced wire-wrapped assemblies

    International Nuclear Information System (INIS)

    Wei, J.P.

    1975-01-01

    The thermal-hydraulic analyses presented in this report are based on application of the subchannel concept in association with the use of bulk parameters for coolant velocity and coolant temperature within a subchannel. The interactions between subchannels are due to turbulent interchange, pressure-induced diversion crossflow, directed sweeping crossflow induced by the helical wire wrap, and transverse thermal conduction. The FULMIX-II computer program was successfully developed to perform the steady-state temperature predictions for LMFBR fuel assemblies with the reference straight-start design and the advanced wire-wrap designs. Predicted steady-state temperature profiles are presented for a typical CRBRP 217-rod wire-wrapped assembly with the selected wire-wrap designs

  13. Handbook of advanced nuclear hydrogen safety. 1st edition

    International Nuclear Information System (INIS)

    Hino, Ryutaro; Takegami, Hiroaki; Ogawa, Toru

    2017-03-01

    In the aftermath of the Fukushima nuclear accident, safety measures against hydrogen in severe accident has been recognized as a serious technical problem in Japan. Therefore, efforts have begun to form a common knowledge base between nuclear engineers and experts on combustion and explosion, and to secure and improve future nuclear energy safety. As one of such activities, we have prepared the 'Handbook of Advanced Nuclear Hydrogen Safety'. A handbook committee consisting of Japanese experts in the fields of nuclear and combustion-explosion in universities, nuclear companies, electric companies and research institutes was established in 2012. The objective and consents of the handbook were determined, and the outline of the contents was decided. The concepts of the handbook are as follows: to show advanced nuclear hydrogen safety technologies that nuclear engineers should understand, to show hydrogen safety points to make combustion-explosion experts cooperate with nuclear engineers, to expand information on water radiolysis considering the situation from just after the Fukushima accidents and to the waste management necessary for decommissioning after the accident etc. Many experts have participated to manuscript preparation, which was the first step of forming a hydrogen community across the boundaries of fields. The hydrogen community is expected to grow along with its improvement to the knowledge base on nuclear hydrogen safety. (author)

  14. Thermal hydraulic feasibility assessment of the spent nuclear fuel project

    International Nuclear Information System (INIS)

    Heard, F.J.

    1996-01-01

    A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The goal was to develop a series of thermal-hydraulic models that could respond to all process and safety related issues that may arise pertaining to the SNFP, as well as provide a basis for validation of the results. Results show that there is a reasonable envelope for process conditions and requirements that are thermally and hydraulically acceptable

  15. Radiation and physical protection challenges at advanced nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Pickett, Susan E.

    2008-01-01

    Full text: The purpose of this study is to examine challenges and opportunities for radiation protection in advanced nuclear reactors and fuel facilities proposed under the Generation IV (GEN IV) initiative which is examining and pursuing the exploration and development of advanced nuclear science and technology; and the Global Nuclear Energy Partnership (GNEP), which seeks to develop worldwide consensus on enabling expanded use of economical, carbon-free nuclear energy to meet growing energy demand. The International Energy Agency projects nuclear power to increase at a rate of 1.3 to 1.5 percent a year over the next 20 years, depending on economic growth. Much of this growth will be in Asia, which, as a whole, currently has plans for 40 new nuclear power plants. Given this increase in demand for new nuclear power facilities, ranging from light water reactors to advanced fuel processing and fabrication facilities, it is necessary for radiation protection and physical protection technologies to keep pace to ensure both worker and public health. This paper is based on a review of current initiatives and the proposed reactors and facilities, primarily the nuclear fuel cycle facilities proposed under the GEN IV and GNEP initiatives. Drawing on the Technology Road map developed under GEN IV, this work examines the potential radiation detection and protection challenges and issues at advanced reactors, including thermal neutron spectrum systems, fast neutron spectrum systems and nuclear fuel recycle facilities. The thermal neutron systems look to improve the efficiency of production of hydrogen or electricity, while the fast neutron systems aim to enable more effective management of actinides through recycling of most components in the discharged fuel. While there are components of these advanced systems that can draw on the current and well-developed radiation protection practices, there will inevitably be opportunities to improve the overall quality of radiation

  16. Advanced display concepts in nuclear control rooms

    International Nuclear Information System (INIS)

    Clark, M.T.; Banks, W.W.; Blackman, H.S.; Gertman, D.I.

    1982-01-01

    Precursors necessary for the development of a full-scale predictor display/control system have been under development since the mid 1940's. The predictor display itself has been available for use in manual control systems since 1958. However, the nuclear industry has not yet explored the uses and benefits of predictor systems. The purpose of this paper is to provide information on the application of this technology to the nuclear industry. The possibility of employing a simulation-based control system for nuclear plant systems that currently use conventional auto/manual schemes is discussed. By employing simulation-based systems, a predictor display could be made available to the operator during manual operations, thus facilitating control without outwardly affecting the overall control scheme

  17. Advancements in nuclear plant maintenance programs

    International Nuclear Information System (INIS)

    Meligi, A.E.; Maras, M.C.

    1993-01-01

    The viability of the nuclear option as a technology choice for present and future electricity generation will be decided primarily on the basis of operating cost to achieve plant performance objectives. In a nuclear plant, performance is judged not only on availability and output rate but also on safety risk and radiation exposure. Operating, cost is essentially made up of the fuel cost and operation and maintenance (O and M) cost. Over the past decade, the industry average nuclear plant performance has improved significantly; however, this improvement was accompanied by rising O and M cost. The net result was that the nuclear option lost its long-standing economic advantage over the coal option, based on the industry average comparison, around 1987 - with the gap narrowing slightly in the last 2 years. In recent times, gas-fired plants have also become a basis for comparison. The electric generation cost comparisons of various fuel options has led to even greater scrutiny of nuclear plant performance, with the poorer performing plants facing the risk of shutdown. While effective O and M programs improve plant performance, present industry data show that there is no direct correlation between the cost of a plant O and M program and its associated performance. There is a significant number of existing tools and techniques in the O and M area that have proved to be successful and have resulted in significant benefits and payback. This paper presents an overview of the nuclear industry efforts to improve the conduct of O and M activities, describes the basic elements of an effective O and M program, and addresses some of the state-of-the-art tools and techniques to enhance maintenance work planning, training, and procedures

  18. Advancing capabilities for detecting undeclared nuclear activities

    International Nuclear Information System (INIS)

    Baute, J.

    2013-01-01

    When a country presents a consistent, transparent and predictable picture of its nuclear programme that is supported by the analysis of all information, IAEA inspectors do not need to go there as frequently for routine verification activities. Rather IAEA can redirect those resources to addressing safeguards issues in the state posing real proliferation concerns. The point is how to establish a coherent picture of a nuclear program and how to identify early warnings of safeguard breaches. A key element is the exploitation of all the information available (open sources, inspection report, satellite imagery, state declarations,...) through effective and quick information analysis. This document is made up of the slides of the presentation

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

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

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

  20. Introducing advanced nuclear fuel cycles in Canada

    International Nuclear Information System (INIS)

    Duret, M.F.

    1978-05-01

    The ability of several different advanced fuel cycles to provide energy for a range of energy growth scenarios has been examined for a few special situations of interest in Canada. Plutonium generated from the CANDU-PHW operating on natural uranium is used to initiate advanced fuel cycles in the year 2000. The four fuel cycles compared are: 1) natural uranium in the CANDU-PHW; 2) high burnup thorium cycle in the CANDU-PHW; 3) self-sufficient thorium cycle in the CANDU-PHW; 4) plutonium-uranium cycle in a fast breeder reactor. The general features of the results are quite clear. While any plutonium generated prior to the introduction of the advanced fuel cycle remains, system requirements for natural uranium for each of the advanced fuel cycles are the same and are governed by the rate at which plants operating on natural uranium can be retired. When the accumulated plutonium inventory has been entirely used, natural uranium is again required to provide inventory for the advanced fuel cycle reactors. The time interval during which no uranium is required varies only from about 25 to 40 years for both thorium cycles, depending primarily on the energy growth rate. The breeder does not require the entire plutonium inventory produced and so would call for less processing of fuel from the PHW reactors. (author)

  1. Advanced construction methods for new nuclear power plants

    International Nuclear Information System (INIS)

    Bilbao y Leon, Sama; Cleveland, John; Moon, Seong-Gyun; Tyobeka, Bismark

    2009-01-01

    The length of the construction and commissioning phases of nuclear power plants have historically been longer than for conventional fossil fuelled plants, often having a record of delays and cost overruns as a result from several factors including legal interventions and revisions of safety regulations. Recent nuclear construction projects however, have shown that long construction periods for nuclear power plants are no longer the norm. While there are several inter-related factors that influence the construction time, the use of advanced construction techniques has contributed significantly to reducing the construction length of recent nuclear projects. (author)

  2. Tutorial on nuclear thermal propulsion safety for Mars

    International Nuclear Information System (INIS)

    Buden, D.

    1992-01-01

    Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments

  3. Advanced nuclear plants: Meeting the economic challenge

    International Nuclear Information System (INIS)

    Redding, J.R.; MacGregor, P.R.

    1993-01-01

    This article examines the economic challenges to nuclear power meeting the forecasted power demand of the end of the century. The topics of the article include the economic challenge, safety, competition from other energy sources, a US case study, environmental costs, capital costs, reducing operation and maintenance costs, and non-technical features

  4. Use of advanced modeling techniques to optimize thermal packaging designs.

    Science.gov (United States)

    Formato, Richard M; Potami, Raffaele; Ahmed, Iftekhar

    2010-01-01

    Through a detailed case study the authors demonstrate, for the first time, the capability of using advanced modeling techniques to correctly simulate the transient temperature response of a convective flow-based thermal shipper design. The objective of this case study was to demonstrate that simulation could be utilized to design a 2-inch-wall polyurethane (PUR) shipper to hold its product box temperature between 2 and 8 °C over the prescribed 96-h summer profile (product box is the portion of the shipper that is occupied by the payload). Results obtained from numerical simulation are in excellent agreement with empirical chamber data (within ±1 °C at all times), and geometrical locations of simulation maximum and minimum temperature match well with the corresponding chamber temperature measurements. Furthermore, a control simulation test case was run (results taken from identical product box locations) to compare the coupled conduction-convection model with a conduction-only model, which to date has been the state-of-the-art method. For the conduction-only simulation, all fluid elements were replaced with "solid" elements of identical size and assigned thermal properties of air. While results from the coupled thermal/fluid model closely correlated with the empirical data (±1 °C), the conduction-only model was unable to correctly capture the payload temperature trends, showing a sizeable error compared to empirical values (ΔT > 6 °C). A modeling technique capable of correctly capturing the thermal behavior of passively refrigerated shippers can be used to quickly evaluate and optimize new packaging designs. Such a capability provides a means to reduce the cost and required design time of shippers while simultaneously improving their performance. Another advantage comes from using thermal modeling (assuming a validated model is available) to predict the temperature distribution in a shipper that is exposed to ambient temperatures which were not bracketed

  5. Advances in Integrated Vehicle Thermal Management and Numerical Simulation

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-10-01

    Full Text Available With the increasing demands for vehicle dynamic performance, economy, safety and comfort, and with ever stricter laws concerning energy conservation and emissions, vehicle power systems are becoming much more complex. To pursue high efficiency and light weight in automobile design, the power system and its vehicle integrated thermal management (VITM system have attracted widespread attention as the major components of modern vehicle technology. Regarding the internal combustion engine vehicle (ICEV, its integrated thermal management (ITM mainly contains internal combustion engine (ICE cooling, turbo-charged cooling, exhaust gas recirculation (EGR cooling, lubrication cooling and air conditioning (AC or heat pump (HP. As for electric vehicles (EVs, the ITM mainly includes battery cooling/preheating, electric machines (EM cooling and AC or HP. With the rational effective and comprehensive control over the mentioned dynamic devices and thermal components, the modern VITM can realize collaborative optimization of multiple thermodynamic processes from the aspect of system integration. Furthermore, the computer-aided calculation and numerical simulation have been the significant design methods, especially for complex VITM. The 1D programming can correlate multi-thermal components and the 3D simulating can develop structuralized and modularized design. Additionally, co-simulations can virtualize simulation of various thermo-hydraulic behaviors under the vehicle transient operational conditions. This article reviews relevant researching work and current advances in the ever broadening field of modern vehicle thermal management (VTM. Based on the systematic summaries of the design methods and applications of ITM, future tasks and proposals are presented. This article aims to promote innovation of ITM, strengthen the precise control and the performance predictable ability, furthermore, to enhance the level of research and development (R&D.

  6. Strategic and policy issues raised by the transition from thermal to fast nuclear systems

    International Nuclear Information System (INIS)

    2009-01-01

    The renewed interest in nuclear energy triggered by concerns about global climate change and security of supply, which could lead to substantial growth in nuclear electricity generation, enhances the attractiveness of fast neutron reactors with closed fuel cycles. Moving from the current fleet of thermal neutron reactors to fast neutron systems will require many decades and extensive RD-D efforts. This book identifies and analyses key strategic and policy issues raised by such a transition, aiming at providing guidance to decision makers on the best approaches for implementing transition scenarios. The topics covered in this book will be of interest to government and nuclear industry policy makers as well as to specialists working on nuclear energy system analyses and advanced fuel cycle issues. (author)

  7. Advances in HELIOS2 nuclear data library

    Energy Technology Data Exchange (ETDEWEB)

    Wemple, Charles [Studsvik Scandpower, Inc., Iadho Falls, ID (United States); Simeonov, Teodosi [Studsvik Scandpower, Inc., Waltham, MA (United States)

    2017-09-15

    The ongoing development of the HELIOS2 code system at Studsvik includes periodic updates of the nuclear data library. The library expansion includes an update of the cross section data source to ENDF/B-VIIR1, a significant expansion of the burnup chains, the addition of a more complete set of gamma production data, and the development of new resonance treatment options. The goal is to provide the capability for HELIOS2 to more accurately model a wider array of reactor applications and enhance interoperability with SNF, the Studsvik spent fuel analysis code. This paper will also provide a discussion of the nuclear data library benchmarking effort and an overview of other HELIOS2 development efforts.

  8. 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

  9. Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

    International Nuclear Information System (INIS)

    2001-05-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  10. Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

    International Nuclear Information System (INIS)

    2001-01-01

    An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for

  11. Assessment of Space Nuclear Thermal Propulsion Facility and Capability Needs

    Energy Technology Data Exchange (ETDEWEB)

    James Werner

    2014-07-01

    The development of a Nuclear Thermal Propulsion (NTP) system rests heavily upon being able to fabricate and demonstrate the performance of a high temperature nuclear fuel as well as demonstrating an integrated system prior to launch. A number of studies have been performed in the past which identified the facilities needed and the capabilities available to meet the needs and requirements identified at that time. Since that time, many facilities and capabilities within the Department of Energy have been removed or decommissioned. This paper provides a brief overview of the anticipated facility needs and identifies some promising concepts to be considered which could support the development of a nuclear thermal propulsion system. Detailed trade studies will need to be performed to support the decision making process.

  12. Some perspectives in nuclear astrophysics on non-thermal phenomena

    International Nuclear Information System (INIS)

    Tatischeff, V.

    2012-01-01

    In this HDR (Accreditation to Supervise Researches) report, the author presents and comments his research activities on nuclear phenomena in stellar eruptions (solar eruptions, lithium nucleosynthesis in stellar eruptions), on particle acceleration in shock waves of stellar explosions (diffusive acceleration by shock wave, particle acceleration in symbiotic novae, particle acceleration in radio-detected supernovae), of research on low energy cosmic rays (galactic emission of nuclear gamma rays, non thermal soft X rays as new tracer of accelerated particles), and on the origin of short period radioactivities in the primitive solar system (extinguished radio-activities and formation of the solar system, origin of berylium-10 in the primitive solar system). The author concludes with some perspectives on non thermal phenomena in nuclear astrophysics, and on research and development for the future of medium-energy gamma astronomy [fr

  13. Advances in Nuclear Power Process Heat Applications

    International Nuclear Information System (INIS)

    2012-05-01

    Following an IAEA coordinated research project, this publication compiles the findings of research and development activities related to practical nuclear process heat applications. An overview of current progress on high temperature gas cooled reactors coupling schemes for different process heat applications, such as hydrogen production and desalination is included. The associated safety aspects are also highlighted. The summary report documents the results and conclusions of the project.

  14. Development of steady thermal-hydraulic analysis code for China advanced research reactor

    International Nuclear Information System (INIS)

    Tian Wenxi; Qiu Suizheng; Guo Yun; Su Guanghui; Jia Dounan; Liu Tiancai; Zhang Jianwei

    2006-01-01

    A multi-channel model steady-state thermal-hydraulic analysis code was developed for China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed flow distribution in the core was obtained. The result shows that the structure size plays the most important role in flow distribution and the influence of core power could be neglected under single-phase flow. The temperature field of fuel element under unsymmetrical cooling condition was also obtained, which is necessary for the further study such as stress analysis etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of hot channel was carried out and it is proved that all thermal-hydraulic parameters accord with the Safety Regulation of CARR. (authors)

  15. A Path Forward to Advanced Nuclear Fuels: Spectroscopic Calorimetry of Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    Tobin, J.G.

    2009-01-01

    The goal is to relieve the shortage of thermodynamic and kinetic information concerning the stability of nuclear fuel alloys. Past studies of the ternary nuclear fuel UPuZr have demonstrated constituent redistribution when irradiated or with thermal treatment. Thermodynamic data is key to predicting the possibilities of effects such as constituent redistribution within the fuel rods and interaction with cladding materials

  16. The status of nuclear fuel cycle system analysis for the development of advanced nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kim, Seong Ki; Lee, Hyo Jik; Chang, Hong Rae; Kwon, Eun Ha; Lee, Yoon Hee; Gao, Fanxing [KAERI, Daejeon (Korea, Republic of)

    2011-11-15

    The system analysis has been used with different system and objectives in various fields. In the nuclear field, the system can be applied from uranium mining to spent fuel reprocessing or disposal which is called the nuclear fuel cycle. The analysis of nuclear fuel cycle can be guideline for development of advanced fuel cycle through integrating and evaluating the technologies. For this purpose, objective approach is essential and modeling and simulation can be useful. In this report, several methods which can be applicable for development of advanced nuclear fuel cycle, such as TRL, simulation and trade analysis were explained with case study

  17. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Thelen, M.C.; Meyer, M.K.; Marshall, F.M.; Foster, J.; Benson, J.B.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  18. Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

    International Nuclear Information System (INIS)

    Schnitzler, Bruce G.

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse (∼900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as

  19. Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06

    International Nuclear Information System (INIS)

    2006-01-01

    Following the highly successful ICAPP'05 meeting held in Seoul Korea, the 2006 International Congress on Advances in Nuclear Power Plants brought together international experts of the nuclear industry involved in the operation, development, building, regulation and research related to Nuclear Power Plants. The program covers the full spectrum of Nuclear Power Plant issues from design, deployment and construction of plants to research and development of future designs and advanced systems. The program covers lessons learned from power, research and demonstration reactors from over 50 years of experience with operation and maintenance, structures, materials, technical specifications, human factors, system design and reliability. The program by technical track deals with: - 1. Water-Cooled Reactor Programs and Issues Evolutionary designs, innovative, passive, light and heavy water cooled reactors; issues related to meeting medium term utility needs; design and regulatory issues; business, political and economic challenges; infrastructure limitations and improved construction techniques including modularization. - 2. High Temperature Gas Cooled Reactors Design and development issues, components and materials, safety, reliability, economics, demonstration plants and environmental issues, fuel design and reliability, power conversion technology, hydrogen production and other industrial uses; advanced thermal and fast reactors. - 3. Long Term Reactor Programs and Strategies Reactor technology with enhanced fuel cycle features for improved resource utilization, waste characteristics, and power conversion capabilities. Potential reactor designs with longer development times such as, super critical water reactors, liquid metal reactors, gaseous and liquid fuel reactors, Gen IV, INPRO, EUR and other programs. - 4. Operation, Performance and Reliability Management Training, O and M costs, life cycle management, risk based maintenance, operational experiences, performance and

  20. Cycle Trades for Nuclear Thermal Rocket Propulsion Systems

    Science.gov (United States)

    White, C.; Guidos, M.; Greene, W.

    2003-01-01

    Nuclear fission has been used as a reliable source for utility power in the United States for decades. Even in the 1940's, long before the United States had a viable space program, the theoretical benefits of nuclear power as applied to space travel were being explored. These benefits include long-life operation and high performance, particularly in the form of vehicle power density, enabling longer-lasting space missions. The configurations for nuclear rocket systems and chemical rocket systems are similar except that a nuclear rocket utilizes a fission reactor as its heat source. This thermal energy can be utilized directly to heat propellants that are then accelerated through a nozzle to generate thrust or it can be used as part of an electricity generation system. The former approach is Nuclear Thermal Propulsion (NTP) and the latter is Nuclear Electric Propulsion (NEP), which is then used to power thruster technologies such as ion thrusters. This paper will explore a number of indirect-NTP engine cycle configurations using assumed performance constraints and requirements, discuss the advantages and disadvantages of each cycle configuration, and present preliminary performance and size results. This paper is intended to lay the groundwork for future efforts in the development of a practical NTP system or a combined NTP/NEP hybrid system.

  1. Advanced Detectors for Nuclear, High Energy and Astroparticle Physics

    CERN Document Server

    Das, Supriya; Ghosh, Sanjay

    2018-01-01

    The book presents high-quality papers presented at a national conference on ‘Advanced Detectors for Nuclear, High Energy and Astroparticle Physics’. The conference was organized to commemorate 100 years of Bose Institute. The book is based on the theme of the conference and provides a clear picture of basics and advancement of detectors for nuclear physics, high-energy physics and astroparticle physics together. The topics covered in the book include detectors for accelerator-based high energy physics; detectors for non-accelerator particle physics; nuclear physics detectors; detection techniques in astroparticle physics and dark matter; and applications and simulations. The book will be a good reference for researchers and industrial personnel working in the area of nuclear and astroparticle physics.

  2. An Overview of Facilities and Capabilities to Support the Development of Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    James Werner; Sam Bhattacharyya; Mike Houts

    2011-02-01

    Abstract. The future of American space exploration depends on the ability to rapidly and economically access locations of interest throughout the solar system. There is a large body of work (both in the US and the Former Soviet Union) that show that Nuclear Thermal Propulsion (NTP) is the most technically mature, advanced propulsion system that can enable this rapid and economical access by its ability to provide a step increase above what is a feasible using a traditional chemical rocket system. For an NTP system to be deployed, the earlier measurements and recent predictions of the performance of the fuel and the reactor system need to be confirmed experimentally prior to launch. Major fuel and reactor system issues to be addressed include fuel performance at temperature, hydrogen compatibility, fission product retention, and restart capability. The prime issue to be addressed for reactor system performance testing involves finding an affordable and environmentally acceptable method to test a range of engine sizes using a combination of nuclear and non-nuclear test facilities. This paper provides an assessment of some of the capabilities and facilities that are available or will be needed to develop and test the nuclear fuel, and reactor components. It will also address briefly options to take advantage of the greatly improvement in computation/simulation and materials processing capabilities that would contribute to making the development of an NTP system more affordable. Keywords: Nuclear Thermal Propulsion (NTP), Fuel fabrication, nuclear testing, test facilities.

  3. maximum neutron flux at thermal nuclear reactors

    International Nuclear Information System (INIS)

    Strugar, P.

    1968-10-01

    Since actual research reactors are technically complicated and expensive facilities it is important to achieve savings by appropriate reactor lattice configurations. There is a number of papers, and practical examples of reactors with central reflector, dealing with spatial distribution of fuel elements which would result in higher neutron flux. Common disadvantage of all the solutions is that the choice of best solution is done starting from the anticipated spatial distributions of fuel elements. The weakness of these approaches is lack of defined optimization criteria. Direct approach is defined as follows: determine the spatial distribution of fuel concentration starting from the condition of maximum neutron flux by fulfilling the thermal constraints. Thus the problem of determining the maximum neutron flux is solving a variational problem which is beyond the possibilities of classical variational calculation. This variational problem has been successfully solved by applying the maximum principle of Pontrjagin. Optimum distribution of fuel concentration was obtained in explicit analytical form. Thus, spatial distribution of the neutron flux and critical dimensions of quite complex reactor system are calculated in a relatively simple way. In addition to the fact that the results are innovative this approach is interesting because of the optimization procedure itself [sr

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

    International Nuclear Information System (INIS)

    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

  5. Advanced plant design recommendations from Cook Nuclear Plant experience

    International Nuclear Information System (INIS)

    Zimmerman, W.L.

    1993-01-01

    A project in the American Electric Power Service Corporation to review operating and maintenance experience at Cook Nuclear Plant to identify recommendations for advanced nuclear plant design is described. Recommendations so gathered in the areas of plant fluid systems, instrument and control, testing and surveillance provisions, plant layout of equipment, provisions to enhance effective maintenance, ventilation systems, radiological protection, and construction, are presented accordingly. An example for a design review checklist for effective plant operations and maintenance is suggested

  6. Recent advances in measurements of the nuclear level density

    International Nuclear Information System (INIS)

    John, Bency

    2007-01-01

    A short review of recent advances in measurements of the nuclear level density is given. First results of the inverse level density parameter - angular momentum correlation in a number of nuclei around Z∼50 shell region at an excitation energy around 0.3 MeV/nucleon are presented. Significant variations observed over and above the expected shell corrections are discussed in context of the emerging trends in microscopic calculations of the nuclear level density. (author)

  7. Advanced Nuclear Reactor Concepts for China

    International Nuclear Information System (INIS)

    Knoche, D.; Sassen, F.; Tietsch, W.; Yujie, Dong; Li, Cao

    2008-01-01

    China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

  8. Advanced Nuclear Reactor Concepts for China

    Energy Technology Data Exchange (ETDEWEB)

    Knoche, D.; Sassen, F.; Tietsch, W. [Westinghouse Electric Germany, Postfach 10 05 63, 68140 Mannheim (Germany); Yujie, Dong; Li, Cao [INET, Tsinghua University, 100084 Beijing (China)

    2008-07-01

    China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

  9. European activities on crosscutting thermal-hydraulic phenomena for innovative nuclear systems

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, X., E-mail: xu.cheng@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Batta, A. [Karlsruhe Institute of Technology (KIT) (Germany); Bandini, G. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Roelofs, F. [Nuclear Research and Consultancy Group (NRG) (Netherlands); Van Tichelen, K. [Studiecentrum voor Kernenergie – Centre d’étude de l’Energie Nucléaire (SCK-CEN) (Belgium); Gerschenfeld, A. [Commissariat à l’Energie Atomique (CEA) (France); Prasser, M. [Paul Scherrer Institute (PSI) (Switzerland); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS) (Germany); Hampel, U. [Helmholtz-Zentrum Dresden-Rossendorf e.V. (HZDR) (Germany); Ma, W.M. [Kungliga Tekniska Högskolan (KTH) (Sweden)

    2015-08-15

    Highlights: • This paper serves as a guidance of the special issue. • The technical tasks and methodologies applied to achieve the objectives have been described. • Main results achieved so far are summarized. - Abstract: Thermal-hydraulics is recognized as a key scientific subject in the development of innovative reactor systems. In Europe, a consortium is established consisting of 24 institutions of universities, research centers and nuclear industries with the main objectives to identify and to perform research activities on important crosscutting thermal-hydraulic issues encountered in various innovative nuclear systems. For this purpose the large-scale integrated research project THINS (Thermal-Hydraulics of Innovative Nuclear Systems) is launched in the 7th Framework Programme FP7 of European Union. The main topics considered in the THINS project are (a) advanced reactor core thermal-hydraulics, (b) single phase mixed convection, (c) single phase turbulence, (d) multiphase flow, and (e) numerical code coupling and qualification. The main objectives of the project are: • Generation of a data base for the development and validation of new models and codes describing the selected crosscutting thermal-hydraulic phenomena. • Development of new physical models and modeling approaches for more accurate description of the crosscutting thermal-hydraulic phenomena. • Improvement of the numerical engineering tools for the design analysis of the innovative nuclear systems. This paper describes the technical tasks and methodologies applied to achieve the objectives. Main results achieved so far are summarized. This paper serves also as a guidance of this special issue.

  10. Trends in the design of advanced nuclear reactors

    International Nuclear Information System (INIS)

    Poong-Eil Juhn; Kupitz, Juergen

    1996-01-01

    Nuclear energy is an essentially unlimited energy source with the potential to provide energy in the form of electricity, district heat and process heat environmentally acceptable conditions. However, this potential will be realized only if nuclear power plants can meet the challenges of national safety requirements, economic competitiveness and public acceptance. Worldwide, a tremendous amount of experience has been accumulated during the development, licensing, construction and operation of nuclear power plants. This experience forms a sound basis for further improvements. Nuclear programmes in the IAEA Member States are addressing the development of advanced reactors, which are intended to have better economics, higher reliability and improved safety. The IAEA, as a global international governmental organization dealing with nuclear power, promotes international information exchange and international co-operation between all countries with their own advanced power programmes and offers assistance to countries with an interest in exploratory or research programmes. The paper gives an overview of global trends in the design of advanced nuclear reactors for electricity generation and heat production along with the role of IAEA. (author)

  11. 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.

  12. Modular construction approach for advanced nuclear plants

    International Nuclear Information System (INIS)

    Johnson, F.T.; Orr, R.S.; Boudreaux, C.P.

    1988-01-01

    Modular construction has been designated as one of the major features of the AP600 program, a small innovative 600-MW (electric) advanced light water reactor (ALWR) that is currently being developed by Westinghouse and its subcontractors. This program is sponsored by the US Department of Energy (DOE) in conjunction with several other DOE and Electric Power Research Institute ALWR programs. Two major objectives of the AP600 program are as follows: (1) to provide a cost of power competitive with other power generation alternatives; and (2) to provide a short construction schedule that can be met with a high degree of certainty. The AP600 plant addresses these objectives by providing a simplified plant design and an optimized plant arrangement that result in a significant reduction in the number and size of systems and components, minimizes the overall building volumes, and consequently reduces the required bulk quantities. However, only by adopting a modular construction approach for the AP600 can the full cost and schedule benefits be realized from the advances made in the plant systems design and plant arrangement. Modularization is instrumental in achieving both of the above objectives, but most of all, a total modularization approach is considered absolutely essential to ensure that an aggressive construction schedule can be met with a high degree of certainty

  13. 11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)

    International Nuclear Information System (INIS)

    Lemonnier, H.

    2005-01-01

    The main topics covered by the NURETH 11 meeting are the thermal-hydraulics of existing and future nuclear power plants as foreseen by the Generation IV worldwide initiative. Normal operation and accidental situations are also relevant topics of the Conference. The topics cover modeling, experiments, instrumentation and numerical simulations related to flow and heat transfer in nuclear reactors with a special emphasis on the advances of multiphase CFD methods. The first part of this Book of Abstracts enumerates the Organizing Scientific Societies, the Sponsors of the Conference, the Conference Chairs, and the members of the Steering Committee and of the Technical Program Committee. The second part of this Book of Abstracts contains the list of the titles of the contributed papers. Each item includes the log number of the paper, the abstract of which can therefore be easily located in the next section of this book. The titles of the papers have been sorted out by topics to provide a synthetic view of the contributions in a selected domain. The last section of this Book includes an index of authors and co-authors with a reference to the log number(s) of their contributed paper(s). Finally, the CD-Rom of the Conference Proceedings containing the full-length papers is inserted at the inside back cover. Sessions content: A - two-phase flow and heat transfer fundamentals: computational and mathematical techniques (numerical schemes, LBM, BEM, mesh-less, etc.); contact angle and wettability phenomena; experiments and data bases for the assessment and the verification of 3D models; flow regime identification and modelling; heat transfer near critical pressure and supercritical water reactors; interfacial area (data base, modeling, measurement techniques); instrumentation techniques; micro-scale basic phenomena, fluid flow and heat transfer; scaling methods; counter current flow; B - code developments: containment analysis; core thermal-hydraulics and subchannel analysis

  14. Advanced waste forms from spent nuclear fuel

    International Nuclear Information System (INIS)

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

    1995-01-01

    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

  15. Recent advances in nuclear power plant simulation

    International Nuclear Information System (INIS)

    Zerbino, H.; Plisson, P.; Friant, J.Y.

    1997-01-01

    The field of industrial simulation has experienced very significant progress in recent years, and power plant simulation in particular has been an extremely active area. Improvements may be recorded in practically all simulator subsystems. In Europe, the construction of new full- or optimized-scope nuclear power plant simulators during the middle 1990's has been remarkable intense. In fact, it is possible to identify a distinct simulator generation, which constitutes a new de facto simulation standard. Thomson Training and Simulation has taken part in these developments by designing, building, and validation several of these new simulators for Dutch, German and French nuclear power plants. Their characteristics are discussed in this paper. The following main trends may be identified: Process modeling is clearly evolving towards obtaining engineering-grade performance, even under the added constraints of real-time operation and a very wide range of operating conditions to be covered; Massive use of modern graphic user interfaces (GUI) ensures an unprecedented flexibility and user-friendliness for the Instructor Station; The massive use of GUIs also allows the development of Trainee Stations (TS), which significantly enhance the in-depth training value of the simulators; The development of powerful Software Development Environments (SDE) enables the simulator maintenance teams to keep abreast of modifications carried out in the reference plants; Finally, simulator maintenance and its compliance with simulator fidelity requirements are greatly enhanced by integrated Configuration Management Systems (CMS). In conclusion, the power plant simulation field has attained a strong level of maturity, which benefits its approximately forty years of service to the power generation industry. (author)

  16. Ground test facilities for evaluating nuclear thermal propulsion engines and fuel elements

    International Nuclear Information System (INIS)

    Allen, G.C.; Beck, D.F.; Harmon, C.D.; Shipers, L.R.

    1992-01-01

    Interagency panels evaluating nuclear thermal propulsion development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and design issues of a proposed ground test complex for evaluating nuclear thermal propulsion engines and fuel elements being developed for the Space Nuclear Thermal Propulsion (SNTP) program. 2 refs

  17. Stochastic modelling of fusion-product transport and thermalization with nuclear elastic scattering

    International Nuclear Information System (INIS)

    Deveaux, J.C.

    1983-01-01

    Monte Carlo methods are developed to model fusion-product (fp) transport and thermalization with both Rutherford scattering and nuclear elastic scattering (NES) in high-temperature (T/sub i/, T/sub e-/ > 50 keV), advanced-fuel (e.g. Cat-D, D- 3 He) plasmas. A discrete-event model is used to superimpose NES collisions on a Rutherford scattering model that contains the Spitzer coefficients of drag, velocity diffusion (VD), and pith-angle scattering (PAS). The effects of NES on fp transport and thermalization are investigated for advanced-fuel, Field-Reversed Mirror (FRM) plasmas that have a significant Hamiltonian-canonical angular momentum (H-Ptheta) space loss cone which scales with the characteristic size (S identical with R/sub HV//3p/sub i/) and applied vacuum magnetic field (B 0 )

  18. 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...

  19. Thermal performance test for steam turbine of nuclear power plants

    International Nuclear Information System (INIS)

    Bu Yubing; Xu Zongfu; Wang Shiyong

    2014-01-01

    Through study of steam turbine thermal performance test of CPR1000 nuclear power plant, we solve the enthalpy calculation problems of the steam turbine in wet steam zone using heat balance method which can help to figure out the real overall heat balance diagram for the first time, and we develop a useful software for thermal heat balance calculation. Ling'ao phase II as an example, this paper includes test instrument layout, system isolation, risk control, data acquisition, wetness measurement, heat balance calculation, etc. (authors)

  20. Interim report on nuclear waste depository thermal analysis

    International Nuclear Information System (INIS)

    Altenbach, T.J.

    1978-01-01

    A thermal analysis of a deep geologic depository for spent nuclear fuel is being conducted. The TRUMP finite difference heat transfer code is used to analyze a 3-dimensional model of the depository. The model uses a unit cell consisting of one spent fuel canister buried in salt beneath a ventilated room in the depository. A base case was studied along with several parametric variations. It is concluded that this method is appropriate for analyzing the thermal response of the system, and that the most important parameter in determining the maximum temperatures is the canister heat generation rate. The effects of room ventilation and different depository media are secondary

  1. Advanced materials for space nuclear power systems

    International Nuclear Information System (INIS)

    Titran, R.H.; Grobstein, T.L.

    1991-01-01

    Research on monolithic refractory metal alloys and on metal matrix composites is being conducted at the NASA Lewis Research Center, Cleveland, Ohio, in support of advanced space power systems. The overall philosophy of the research is to develop and characterize new high-temperature power conversion and radiator materials and to provide spacecraft designers with material selection options and design information. Research on three candidate materials (carbide strengthened niobium alloy PWC-11 for fuel cladding, graphite fiber reinforced copper matrix composites (Gr/Cu) for heat rejection fins, and tungsten fiber reinforced niobium matrix composites (W/NB) for fuel containment and structural supports) considered for space power system applications is discussed. Each of these types of materials offers unique advantages for space power applications

  2. Development of advanced nuclear reactors in Russia

    International Nuclear Information System (INIS)

    Sotoudeh, M.; Silakhori, K.; Sepanloo, K.; Jahanfarnia, G.; Moattar, F.

    2008-01-01

    Several advanced reactor designs have been so far developed in Russia. The AES-91 and AES-92 plants with the VVER-1000 reactors have been developed at the beginning of 1990. However, the former design has been built in China and the latest which is certified meeting European Utility Requirements is being built in India. Moreover, the model VVER-1500 reactor with 50-60 MWd/t burn-up and an enhanced safety was being developed by Gidropress about 2005, excepting to be completed in 2007. But, this schedule has slipped in favor of development of the AES-2006 power plant incorporating a third-generation standardized VVER-1200 reactor of 1170 MWe. This is an evolutionary development of the well-proven VVER-1000 reactor in the AES-92 plant, with longer life, greater power and efficiency and its lead units are being built at Novovoronezh II, to start operation in 2012-13. Based on Atomenergoproekt declaration, the AES-2006 conforms to both Russian standards and European Utility Requirements. The most important features of the AES-2006 design are mentioned as: a design based on the passive safety systems, double containment, longer plant service life of 50 years with a capacity factor of 92%, longer irreplaceable components service life of 60 years, a 28.6% lower amount of concrete and metal, shorter construction time of 54 months, a Core Damage Frequency of 1x10 -7 / year and lower liquid and solid wastes by 70% and 80% respectively. The presented paper includes a comparative analysis of technological and safety features, economic parameters and environmental impact of the AES-2006 design versus the other western advanced reactors. Since the Bushehr phase II NPP and several other NPPs are planning in Iran, such analysis would be of a great importance

  3. Thermal properties of nuclear matter under the periodic boundary condition

    International Nuclear Information System (INIS)

    Otuka, Naohiko; Ohnishi, Akira

    1999-01-01

    We present the thermal properties of nuclear matter under the periodic boundary condition by the use of our hadronic nucleus-nucleus cascade model (HANDEL) which is developed to treat relativistic heavy-ion collisions from BNL-AGS to CERN-SPS. We first show some results of p-p scattering calculation in our new version which is improved in order to treat isospin ratio and multiplicity more accurately. We then display the results of calculation of nuclear matter with baryon density ρ b = 0.77 fm 3 at some energy densities. Time evolution of particle abundance and temperature are shown. (author)

  4. Grooved Fuel Rings for Nuclear Thermal Rocket Engines

    Science.gov (United States)

    Emrich, William

    2009-01-01

    An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

  5. Nuclear thermal rocket workshop reference system Rover/NERVA

    International Nuclear Information System (INIS)

    Borowski, S.K.

    1991-01-01

    The Rover/NERVA engine system is to be used as a reference, against which each of the other concepts presented in the workshop will be compared. The following topics are reviewed: the operational characteristics of the nuclear thermal rocket (NTR); the accomplishments of the Rover/NERVA programs; and performance characteristics of the NERVA-type systems for both Mars and lunar mission applications. Also, the issues of ground testing, NTR safety, NASA's nuclear propulsion project plans, and NTR development cost estimates are briefly discussed

  6. Thermal effluents from nuclear power plant influences species distribution and thermal tolerance of fishes in reservoirs

    International Nuclear Information System (INIS)

    Pal, A.K.; Das, T.; Dalvi, R.S.; Bagchi, S.; Manush, S.M.; Ayyappan, S.; Chandrachoodan, P.P.; Apte, S.K.; Ravi, P.M.

    2007-01-01

    During electricity generation water bodies like reservoir act as a heat sink for thermal effluent discharges from nuclear power plant. We hypothesized that the fish fauna gets distributed according to their temperature preference in the thermal gradient. In a simulated environment using critical thermal methodology (CTM), we assessed thermal tolerance and metabolic profile of fishes (Puntius filamentosus, Parluciosoma daniconius, Ompok malabaricus, Mastacembelus armatus, Labeo calbasu, Horabragrus brachysoma, Etroplus suratensis, Danio aequipinnatus and Gonoproktopterus curmuca) collected from Kadra reservoir in Karnataka state. Results of CTM tests agrees with the species abundance as per the temperature gradient formed in the reservoir due to thermal effluent discharge. E. suratensis and H. brachysoma) appear to be adapted to high temperature (with high CTMax and CTMin values) and are in abundance at point of thermal discharge. Similarly, P. daniconius, appear to be adapted to cold (low CTM values) is in abundance in lower stretches of Kadra reservoir. Overall results indicate that discharge form nuclear power plant influences the species biodiversity in enclosed water bodies. (author)

  7. Development of vacuum glazing with advanced thermal properties - Final report

    Energy Technology Data Exchange (ETDEWEB)

    Koebel, M.; Manz, H.

    2009-03-15

    Windows constitute a weak link in the building envelope and hence contribute significantly to the total heating energy demand in buildings. By evacuating the glazing cavity a vacuum glazing is created and heat transfer can be significantly reduced. This project was designed to build knowledge and technology necessary to fabricate vacuum glazing with advanced thermal properties. More specifically, various strategies for improvement of conventional technology were investigated. Of central importance was the development of a novel edge sealing approach which can in theory circumvent the main limitation of conventional glass soldering technology. This approach which is rapid, low temperature, low cost and completely vacuum compatible was filed for patenting in 2008. With regards to thermal insulation performance and glazing deflection, numerical studies were performed demonstrating the importance of nonlinear behavior with glazing size and the results published. A detailed service life prediction model was elaborated which defines a set of parameters necessary to keep the expected pressure increase below a threshold value of 0.1 Pa after 30 years. The model takes into account four possible sources of pressure increase and a getter material which acts as a sink. For the production of 0.5 m by 0.5 m glazing assembly prototypes, a high vacuum chamber was constructed and a first sealing prototype realized therein. The manufacture of improved prototypes and optimization of the anodic bonding edge sealing technology with emphasis on process relevant aspects is the goal of a follow-up project. (authors)

  8. Cost estimate guidelines for advanced nuclear power technologies

    International Nuclear Information System (INIS)

    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

  9. Cost estimate guidelines for advanced nuclear power technologies

    International Nuclear Information System (INIS)

    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

  10. Cost estimate guidelines for advanced nuclear power technologies

    International Nuclear Information System (INIS)

    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

  11. Nuclear thermal rocket nozzle testing and evaluation program

    International Nuclear Information System (INIS)

    Davidian, K.O.; Kacynski, K.J.

    1993-01-01

    Performance characteristics of the Nuclear Thermal Rocket can be enhanced through the use of unconventional nozzles as part of the propulsion system. In this report, the Nuclear Thermal Rocket nozzle testing and evaluation program being conducted at the NASA Lewis Research Center is outlined and the advantages of a plug nozzle are described. A facility description, experimental designs and schematics are given. Results of pretest performance analyses show that high nozzle performance can be attained despite substantial nozzle length reduction through the use of plug nozzles as compared to a convergent-divergent nozzle. Pretest measurement uncertainty analyses indicate that specific impulse values are expected to be within plus or minus 1.17%

  12. Spent nuclear fuel storage pool thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Gay, R.R.

    1984-01-01

    Storage methods and requirements for spent nuclear fuel at U.S. commercial light water reactors are reviewed in Section 1. Methods of increasing current at-reactor storage capabilities are also outlined. In Section 2 the development of analytical methods for the thermal-hydraulic analysis of spent fuel pools is chronicled, leading up to a discussion of the GFLOW code which is described in Section 3. In Section 4 the verification of GFLOW by comparisons of the code's predictions to experimental data taken inside the fuel storage pool at the Maine Yankee nuclear power plant is presented. The predictions of GFLOW using 72, 224, and 1584 node models of the storage pool are compared to each other and to the experimental data. An example of thermal licensing analysis for Maine Yankee using the GFLOW code is given in Section 5. The GFLOW licensing analysis is compared to previous licensing analysis performed by Yankee Atomic using the RELAP-4 computer code

  13. Advanced I and C systems for nuclear power plants

    International Nuclear Information System (INIS)

    Bock, H.W.; Graf, A.; Hofmann, H.

    1995-01-01

    Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally, specific requirements coming from the nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS. The main features of these systems are the clear task related architecture with adaptable redundancy, the consequent application of standards for interfaces and communication, comprehensive tools for easy design and service and a highly ergonomic screen based man-machine-interface. The engineering tasks are supported by an integrated engineering system, which has the capacity for design, test and diagnosis of all I and C functions and the related equipment. TELEPERM XP is designed to optimally perform all automatic functions, which require no nuclear specific qualification. This includes all sequences and closed-loop controls as well as most man-machine-interface functions. TELEPERM XS is designed for all control tasks which require a nuclear specific qualification. This especially includes all functions to initiate automatic countermeasures to prevent or to cope with accidents. By use of the complementary I and C systems TELEPERM XP and TELEPERM XS, economical as well as advanced plant automation and man-machine-interfaces can be implemented into Nuclear Plants, assuring the compliance with the relevant international safety standards. (author). 10 figs

  14. Advanced I and C systems for nuclear power plants

    International Nuclear Information System (INIS)

    Bock, H.W.; Graf, A.; Hofmann, H.

    1997-01-01

    Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally specific requirements arising from nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS. The main features of these systems are a clear task related architecture with adaptable redundancy, a consequent application of standards for interfaces and communication, comprehensive tools for easy design and service and a highly ergonomic screen based man-machine-interface. The engineering tasks are supported by an integrated engineering system, which has the capacity for design, test and diagnosis of all I and C functions and the related equipment. TELEPERM XP is designed to optimally perform all automatic functions, which require no nuclear specific qualification. This includes all sequences and closed-loop controls as well as most man-machine-interface functions. TELEPERM XS is designed for all control tasks which require a nuclear specific qualification. This especially includes all function to initiated automatic countermeasures to prevent or to cope with accidents. By use of the complementary I and C systems TELEPERM XP and TELEPERM XS, advanced and likewise economical plant automation and man-machine-interfaces can be implemented into Nuclear Power Plant, assuring compliance with the relevant international safety standards. (author). 10 figs

  15. Advanced I and C systems for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bock, H W; Graf, A; Hofmann, H [Siemens AG, Erlangen (Germany)

    1997-07-01

    Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally specific requirements arising from nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS. The main features of these systems are a clear task related architecture with adaptable redundancy, a consequent application of standards for interfaces and communication, comprehensive tools for easy design and service and a highly ergonomic screen based man-machine-interface. The engineering tasks are supported by an integrated engineering system, which has the capacity for design, test and diagnosis of all I and C functions and the related equipment. TELEPERM XP is designed to optimally perform all automatic functions, which require no nuclear specific qualification. This includes all sequences and closed-loop controls as well as most man-machine-interface functions. TELEPERM XS is designed for all control tasks which require a nuclear specific qualification. This especially includes all function to initiated automatic countermeasures to prevent or to cope with accidents. By use of the complementary I and C systems TELEPERM XP and TELEPERM XS, advanced and likewise economical plant automation and man-machine-interfaces can be implemented into Nuclear Power Plant, assuring compliance with the relevant international safety standards. (author). 10 figs.

  16. Advanced teleoperation in nuclear applications: consolidated fuel reprocessing program

    International Nuclear Information System (INIS)

    Hamel, W.R.; Feldman, M.J.; Martin, H.L.

    1984-01-01

    A new generation of integrated remote maintenance systems is being developed to meet the needs of future nuclear fuel reprocessing at the Oak Ridge National Laboratory. Development activities cover all aspects of an advanced teleoperated maintenance system with particular emphasis on a new force-reflecting servomanipulator concept. The new manipulator, called the advanced servomanipulator, is microprocessor controlled and is designed to achieve force-reflection performance near that of mechanical master/slave manipulators. The advanced servomanipulator uses a gear-drive transmission which permits modularization for remote maintainability (by other advanced servomanipulators) and increases reliability. Human factors analysis has been used to develop an improved man/machine interface concept based upon colographic displays and menu-driven touch screens. Initial test and evaluation of two advanced servomanipulator slave arms and several other development components have begun. 9 references, 5 figures

  17. Thermohydraulic modeling of nuclear thermal rockets: The KLAXON code

    International Nuclear Information System (INIS)

    Hall, M.L.; Rider, W.J.; Cappiello, M.W.

    1992-01-01

    The hydrogen flow from the storage tanks, through the reactor core, and out the nozzle of a Nuclear Thermal Rocket is an integral design consideration. To provide an analysis and design tool for this phenomenon, the KLAXON code is being developed. A shock-capturing numerical methodology is used to model the gas flow (the Harten, Lax, and van Leer method, as implemented by Einfeldt). Preliminary results of modeling the flow through the reactor core and nozzle are given in this paper

  18. NERVA-Derived Concept for a Bimodal Nuclear Thermal Rocket

    International Nuclear Information System (INIS)

    Fusselman, Steven P.; Frye, Patrick E.; Gunn, Stanley V.; Morrison, Calvin Q.; Borowski, Stanley K.

    2005-01-01

    The Nuclear Thermal Rocket is an enabling technology for human exploration missions. The 'bimodal' NTR (BNTR) provides a novel approach to meeting both propulsion and power requirements of future manned and robotic missions. The purpose of this study was to evaluate tie-tube cooling configurations, NTR performance, Brayton cycle performance, and LOX-Augmented NTR (LANTR) feasibility to arrive at a point of departure BNTR configuration for subsequent system definition

  19. Identifying and addressing specific student difficulties in advanced thermal physics

    Science.gov (United States)

    Smith, Trevor I.

    As part of an ongoing multi-university research study on student understanding of concepts in thermal physics at the upper division, I identified several student difficulties with topics related to heat engines (especially the Carnot cycle), as well as difficulties related to the Boltzmann factor. In an effort to address these difficulties, I developed two guided-inquiry worksheet activities (a.k.a. tutorials) for use in advanced undergraduate thermal physics courses. Both tutorials seek to improve student understanding of the utility and physical background of a particular mathematical expression. One tutorial focuses on a derivation of Carnot's theorem regarding the limit on thermodynamic efficiency, starting from the Second Law of Thermodynamics. The other tutorial helps students gain an appreciation for the origin of the Boltzmann factor and when it is applicable; focusing on the physical justification of its mathematical derivation, with emphasis on the connections between probability, multiplicity, entropy, and energy. Student understanding of the use and physical implications of Carnot's theorem and the Boltzmann factor was assessed using written surveys both before and after tutorial instruction within the advanced thermal physics courses at the University of Maine and at other institutions. Classroom tutorial sessions at the University of Maine were videotaped to allow in-depth scrutiny of student successes and failures following tutorial prompts. I also interviewed students on various topics related to the Boltzmann factor to gain a more complete picture of their understanding and inform tutorial revisions. Results from several implementations of my tutorials at the University of Maine indicate that students did not have a robust understanding of these physical principles after lectures alone, and that they gain a better understanding of relevant topics after tutorial instruction; Fisher's exact tests yield statistically significant improvement at the

  20. Nuclear power plant thermal-hydraulic performance research program plan

    International Nuclear Information System (INIS)

    1988-07-01

    The purpose of this program plan is to present a more detailed description of the thermal-hydraulic research program than that provided in the NRC Five-Year Plan so that the research plan and objectives can be better understood and evaluated by the offices concerned. The plan is prepared by the Office of Nuclear Regulatory Research (RES) with input from the Office of Nuclear Reactor Regulation (NRR) and updated periodically. The plan covers the research sponsored by the Reactor and Plant Systems Branch and defines the major issues (related to thermal-hydraulic behavior in nuclear power plants) the NRC is seeking to resolve and provides plans for their resolution; relates the proposed research to these issues; defines the products needed to resolve these issues; provides a context that shows both the historical perspective and the relationship of individual projects to the overall objectives; and defines major interfaces with other disciplines (e.g., structural, risk, human factors, accident management, severe accident) needed for total resolution of some issues. This plan addresses the types of thermal-hydraulic transients that are normally considered in the regulatory process of licensing the current generation of light water reactors. This process is influenced by the regulatory requirements imposed by NRC and the consequent need for technical information that is supplied by RES through its contractors. Thus, most contractor programmatic work is administered by RES. Regulatory requirements involve the normal review of industry analyses of design basis accidents, as well as the understanding of abnormal occurrences in operating reactors. Since such transients often involve complex thermal-hydraulic interactions, a well-planned thermal-hydraulic research plan is needed

  1. Nuclear data for the calculation of thermal reactor reactivity coefficients

    International Nuclear Information System (INIS)

    1989-01-01

    On its 15th meeting in Vienna, 16-20 June 1986, the International Nuclear Data Committee (INDC) considered it important to review the accuracy with which changes in thermal reactor reactivity resulting from changes in temperature and coolant density can be predicted. It was noted that reactor physicists in several countries had to adjust the thermal neutron cross-section data base in order to reproduce measured reactivity coefficients. Consequently, it appeared to be essential to examine the consistency of the integral and differential cross-section data and to make all the information available which has a bearing on reactivity coefficient prediction. Following the recommendation of the INDC, the Nuclear Data Section of the International Atomic Energy Agency, therefore, convened the Advisory Group Meeting on Nuclear Data for the Calculation of Thermal Reaction Reactivity Coefficients, in Vienna, Austria, 7-10 Dec. 1987. The Conclusions and Recommendations of the meeting together with the papers presented, are submitted in the present document. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs

  2. Nuclear Explosion Monitoring Advances and Challenges

    Science.gov (United States)

    Baker, G. E.

    2015-12-01

    We address the state-of-the-art in areas important to monitoring, current challenges, specific efforts that illustrate approaches addressing shortcomings in capabilities, and additional approaches that might be helpful. The exponential increase in the number of events that must be screened as magnitude thresholds decrease presents one of the greatest challenges. Ongoing efforts to exploit repeat seismic events using waveform correlation, subspace methods, and empirical matched field processing holds as much "game-changing" promise as anything being done, and further efforts to develop and apply such methods efficiently are critical. Greater accuracy of travel time, signal loss, and full waveform predictions are still needed to better locate and discriminate seismic events. Important developments include methods to model velocities using multiple types of data; to model attenuation with better separation of source, path, and site effects; and to model focusing and defocusing of surface waves. Current efforts to model higher frequency full waveforms are likely to improve source characterization while more effective estimation of attenuation from ambient noise holds promise for filling in gaps. Censoring in attenuation modeling is a critical problem to address. Quantifying uncertainty of discriminants is key to their operational use. Efforts to do so for moment tensor (MT) inversion are particularly important, and fundamental progress on the statistics of MT distributions is the most important advance needed in the near term in this area. Source physics is seeing great progress through theoretical, experimental, and simulation studies. The biggest need is to accurately predict the effects of source conditions on seismic generation. Uniqueness is the challenge here. Progress will depend on studies that probe what distinguishes mechanisms, rather than whether one of many possible mechanisms is consistent with some set of observations.

  3. Thermal hydraulic-Mechanic Integrated Simulation for Advanced Cladding Thermal Shock Fracture Analysis during Reflood Phase in LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seong Min; Lee, You Ho; Cho, Jae Wan; Lee, Jeong Ik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This study suggested thermal hydraulic-mechanical integrated stress based methodology for analyzing the behavior of ATF type claddings by SiC-Duplex cladding LBLOCA simulation. Also, this paper showed that this methodology could predict real experimental result well. That concept for enhanced safety of LWR called Advanced Accident-Tolerance Fuel Cladding (ATF cladding, ATF) is researched actively. However, current nuclear fuel cladding design criteria for zircaloy cannot be apply to ATF directly because those criteria are mainly based on limiting their oxidation. So, the new methodology for ATF design criteria is necessary. In this study, stress based analysis methodology for ATF cladding design criteria is suggested. By simulating LBLOCA scenario of SiC cladding which is the one of the most promising candidate of ATF. Also we'll confirm our result briefly through comparing some facts from other experiments. This result is validating now. Some of results show good performance with 1-D failure analysis code for SiC fuel cladding that already developed and validated by Lee et al,. It will present in meeting. Furthermore, this simulation presented the possibility of understanding the behavior of cladding deeper. If designer can predict the dangerous region and the time precisely, it may be helpful for designing nuclear fuel cladding geometry and set safety criteria.

  4. Main research results in the field of nuclear power engineering of the Nuclear Reactors and Thermal Physics Institute in 2014

    International Nuclear Information System (INIS)

    Trufanov, A.A.; Orlov, Yu.I.; Sorokin, A.P.; Chernonog, V.L.

    2015-01-01

    The main results of scientific and technological activities for last years of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in solving problems of nuclear power engineering are presented. The work have been carried out on the following problems: justification of research and development solutions and safety of NPPs with fast reactors of new generation with sodium (BN-1200, MBIR) and lead (BREST-OD-300) coolants, justification of safety of operating and advanced NPPs with WWER reactor facilities (WWER-1000, AEhS 2006, WWER-TOI), development and benchmarking of computational codes, research and development support of Beloyarsk-3 (BN-600) and Bilibino (BN-800) NPPs operation, decommissioning of AM and BR-10 research reactors, pilot scientific studies (WWER-SKD, ITER), international scientific and technical cooperation. Problems for further investigations are charted [ru

  5. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

    This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

  6. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 2, Sessions 6-11

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

    This document, Volume 2, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

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

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  9. A Novel Method To On-Line Monitor Reactor Nuclear Power And In-Core Thermal Environments

    International Nuclear Information System (INIS)

    Liu, Hanying; Miller, Don W.; Li, Dongxu; Radcliff, Thomas D.

    2002-01-01

    For current nuclear power plants, nuclear power can not be directly measured and in-core fuel thermal environments can not be monitored due to the unavailability of an appropriate measurement technology and the inaccessibility of the fuel. If the nuclear deposited power and the in-core thermal conditions (i.e. fuel or coolant temperature and heat transfer coefficient) can be monitored in-situ, then it would play a valuable and critical role in increasing nuclear power, predicting abnormal reactor operation, improving core physical models and reducing core thermal margin so as to implement higher fuel burn-up. Furthermore, the management of core thermal margin and fuel operation may be easier during reactor operation, post-accident or spent fuel storage. On the other hand, for some advanced Generation IV reactors, the sealed and long-lived reactor core design challenges traditional measurement techniques while conventional ex-core detectors and current in-core detectors can not monitor details of the in-core fuel conditions. A method is introduced in this paper that responds to the challenge to measure nuclear power and to monitor the in-core thermal environments, for example, local fuel pin or coolant heat convection coefficient and temperature. In summary, the method, which has been designed for online in-core measurement and surveillance, will be beneficial to advanced plant safety, efficiency and economics by decreasing thermal margin or increasing nuclear power. The method was originally developed for a constant temperature power sensor (CTPS). The CTPS is undergoing design and development for an advanced reactor core to measure in-core nuclear power in measurement mode and to monitor thermal environments in compensation mode. The sensor dynamics was analyzed in compensation mode to determine the environmental temperature and the heat transfer coefficient. Previous research demonstrated that a first order dynamic model is not sufficient to simulate sensor

  10. Thermal and statistical properties of nuclei and nuclear systems

    International Nuclear Information System (INIS)

    Moretto, L.G.; Wozniak, G.J.

    1989-07-01

    The term statistical decay, statistical or thermodynamic equilibrium, thermalization, temperature, etc., have been used in nuclear physics since the introduction of the compound nucleus (CN) concept, and they are still used, perhaps even more frequently, in the context of intermediate- and high-energy heavy-ion reactions. Unfortunately, the increased popularity of these terms has not made them any clearer, and more often than not one encounters sweeping statements about the alleged statisticity of a nuclear process where the ''statistical'' connotation is a more apt description of the state of the speaker's mind than of the nuclear reaction. It is our goal, in this short set of lectures, to set at least some ideas straight on this broad and beautiful subject, on the one hand by clarifying some fundamental concepts, on the other by presenting some interesting applications to actual physical cases. 74 refs., 38 figs

  11. Design considerations for Mars transfer vehicles using nuclear thermal propulsion

    Science.gov (United States)

    Emrich, William J.

    1995-01-01

    The design of a Mars Transfer Vehicle (MTV) utilizing nuclear propulsion will require that careful consideration be given to the nuclear radiation environment in which it will operate. The extremely high neutron and gamma fluxes characteristic of nuclear thermal propulsion systems will cause significant heating of the fluid systems in close proximity to the reactor, especially in the lower propellant tanks. Crew radiation doses are also a concern particularly late in a mission when there is less shielding from the propellant tanks. In this study, various vehicle configuration and shielding strategies were examined and the resulting time dependent radiation fields evaluated. A common cluster of three particle bed reactor (PBR) engines were used in all configurations examined. In general, it appears that long, relatively narrow vehicles perform the best from a radiation standpoint, however, good shield optimization will be critical in maintaining a low radiation environment while minimizing the shield weight penalty.

  12. Proceedings of GLOBAL 2007 conference on advanced nuclear fuel cycles and systems

    International Nuclear Information System (INIS)

    2007-01-01

    In keeping with the 12-year history of this conference, GLOBAL 2007 focuses on future nuclear energy systems and fuel cycles. With the increasing public acceptance and political endorsement of nuclear energy, it is a pivotal time for nuclear energy research. Significant advances have been made in development of advanced nuclear fuels and materials, reactor designs, partitioning, transmutation and reprocessing technologies, and waste management strategies. In concert with the technological advances, it is more important than ever to develop sensible nuclear proliferation policies, to promote sustainability, and to continue to increase international collaboration. To further these aims, GLOBAL 2007 highlights recent developments in the following areas: advanced integrated fuel cycle concepts, spent nuclear fuel reprocessing, advanced reprocessing technology, advanced fuels and materials, advanced waste management technology, novel concepts for waste disposal and repository development, advanced reactors, partitioning and transmutation, developments in nuclear non-proliferation technology, policy, and implementation, sustainability and expanded global utilization of nuclear energy, and international collaboration on nuclear energy

  13. Transmutation of Thermocouples in Thermal and Fast Nuclear Reactors

    International Nuclear Information System (INIS)

    Scervini, M.; Rae, C.; Lindley, B.

    2013-06-01

    Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. Their role is fundamental for the control of current nuclear reactors and for the development of the nuclear technology needed for the implementation of GEN IV nuclear reactors. When used for in-core measurements thermocouples are strongly affected not only by high temperatures, but also by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition in the thermoelements and, as a consequence, a time dependent drift in the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation for relatively short times compared to the life required for temperature sensors in nuclear applications. In this work, undertaken as part of the European project METROFISSION, the change in composition occurring in irradiated thermocouples has been calculated using the software ORIGEN 2.2. Several thermocouples have been considered, including Nickel based thermocouples (type K and type N), Tungsten based thermocouples (W-5%Re vs W-26%Re and W- 3%Re vs W-25%Re), Platinum based thermocouples (type S and Platinum vs Palladium) and Molybdenum vs Niobium thermocouples. The transmutation induced by both thermal flux and fast flux has been calculated. Thermocouples undergo more pronounced transmutation in thermal fluxes rather than in fast fluxes, as the neutron cross section of an element is higher for thermal energies. Nickel based thermocouples have a minimal change in composition, while Platinum based and Tungsten based thermocouples experience a very significant transmutation. The use of coatings deposited on the sheath of a thermocouple has been considered as a mean to reduce the neutron flux the thermoelements inside the thermocouple sheath

  14. Space Nuclear Thermal Propulsion Test Facilities Subpanel. Final report

    International Nuclear Information System (INIS)

    Allen, G.C.; Warren, J.W.; Martinell, J.; Clark, J.S.; Perkins, D.

    1993-04-01

    On 20 Jul. 1989, in commemoration of the 20th anniversary of the Apollo 11 lunar landing, President George Bush proclaimed his vision for manned space exploration. He stated, 'First for the coming decade, for the 1990's, Space Station Freedom, the next critical step in our space endeavors. And next, for the new century, back to the Moon. Back to the future. And this time, back to stay. And then, a journey into tomorrow, a journey to another planet, a manned mission to Mars.' On 2 Nov. 1989, the President approved a national space policy reaffirming the long range goal of the civil space program: to 'expand human presence and activity beyond Earth orbit into the solar system.' And on 11 May 1990, he specified the goal of landing Astronauts on Mars by 2019, the 50th anniversary of man's first steps on the Moon. To safely and ever permanently venture beyond near Earth environment as charged by the President, mankind must bring to bear extensive new technologies. These include heavy lift launch capability from Earth to low-Earth orbit, automated space rendezvous and docking of large masses, zero gravity countermeasures, and closed loop life support systems. One technology enhancing, and perhaps enabling, the piloted Mars missions is nuclear propulsion, with great benefits over chemical propulsion. Asserting the potential benefits of nuclear propulsion, NASA has sponsored workshops in Nuclear Electric Propulsion and Nuclear Thermal Propulsion and has initiated a tri-agency planning process to ensure that appropriate resources are engaged to meet this exciting technical challenge. At the core of this planning process, NASA, DOE, and DOD established six Nuclear Propulsion Technical Panels in 1991 to provide groundwork for a possible tri-agency Nuclear Propulsion Program and to address the President's vision by advocating an aggressive program in nuclear propulsion. To this end the Nuclear Electric Propulsion Technology Panel has focused it energies

  15. RECENT ACTIVITIES AT THE CENTER FOR SPACE NUCLEAR RESEARCH FOR DEVELOPING NUCLEAR THERMAL ROCKETS

    International Nuclear Information System (INIS)

    O'Brien, Robert C.

    2001-01-01

    Nuclear power has been considered for space applications since the 1960s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors/ rocket-engines in the Rover/NERVA programs. However, changes in environmental laws may make the redevelopment of the nuclear rocket more difficult. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel form significantly different from NERVA may be needed to ensure public support. The Center for Space Nuclear Research (CSNR) is pursuing development of tungsten based fuels for use in a NTR, for a surface power reactor, and to encapsulate radioisotope power sources. The CSNR Summer Fellows program has investigated the feasibility of several missions enabled by the NTR. The potential mission benefits of a nuclear rocket, historical achievements of the previous programs, and recent investigations into alternatives in design and materials for future systems will be discussed.

  16. Advanced applications of water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2008-07-01

    By August 2007, there were 438 nuclear power plants (NPPs) in operation worldwide, with a total capacity of 371.7 GW(e). Further, 31 units, totaling 24.1 GW(e), were under construction. During 2006 nuclear power produced 2659.7 billion kWh of electricity, which was 15.2% of the world's total. The vast majority of these plants use water-cooled reactors. Based on information provided by its Member States, the IAEA projects that nuclear power will grow significantly, producing between 2760 and 2810 billion kWh annually by 2010, between 3120 and 3840 billion kWh annually by 2020, and between 3325 and 5040 billion kWh annually by 2030. There are several reasons for these rising expectations for nuclear power: - Nuclear power's lengthening experience and good performance: The industry now has more than 12 000 reactor years of experience, and the global average nuclear plant availability during 2006 reached 83%; - Growing energy needs: All forecasts project increases in world energy demand, especially as population and economic productivity grow. The strategies are country dependent, but usually involve a mix of energy sources; - Interest in advanced applications of nuclear energy, such as seawater desalination, steam for heavy oil recovery and heat and electricity for hydrogen production; - Environmental concerns and constraints: The Kyoto Protocol has been in force since February 2005, and for many countries (most OECD countries, the Russian Federation, the Baltics and some countries of the Former Soviet Union and Eastern Europe) greenhouse gas emission limits are imposed; - Security of energy supply is a national priority in essentially every country; and - Nuclear power is economically competitive and provides stability of electricity price. In the near term most new nuclear plants will be evolutionary water cooled reactors (Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs), often pursuing economies of scale. In the longer term, innovative designs that

  17. Advanced nuclear turbojet powerplant characteristics summary for supersonic aircraft

    International Nuclear Information System (INIS)

    Larson, John W.

    1959-01-01

    The estimated powerplant characteristics of an advanced nuclear powerplant intended for use in a nuclear supersonic manned airplane is contained in this report. This nuclear powerplant consists of a 575 MW, high temperature, lithium-cooled, solid fuel element-type reactor coupled to six turbojet engines especially designed for a supersonic nuclear airplane. The lithium coolant passes from the reactor at 2000F directly to the engine radiators without the use of an intermediate heat exchanger. The engines are fitted with burners enabling the thrust produced by the nuclear powerplant to be augmented by the use of chemical fuel for the take-off, transonic acceleration and landing portions of the flight. The powerplant components have been selected for a maximum thrust-to-weight ratio at Mach 3 and 55,000 feet altitude on nuclear heat only operation compromised for net thrust produced with chemical fuel augmentation during the transonic portion of flight. The power plant data presented, therefore, are primarily applicable to an all supersonic mission on nuclear heat alone. The powerplant data presented in this report are an extension of data contained in PWAC-243, 'NJ-14 All-Nuclear Supersonic Bomber Powerplant Characteristics Summary, March 11, 1958', to a higher reactor power. In addition, the engine compressor pressure ratio has been increased to improve transonic thrust characteristics. Weight data are tabulated for the 575 MW powerplant. The engine envelope based on preliminary radiator size estimates is illustrated. A liquid metal system flow schematic and piping data are included. Shield information including reactor shield outline, assumptions, weights, and direct dose pattern at 50 feet is also included. Estimated performance on nuclear heat only operation and nuclear heat plus burning is presented for an envelope of flight conditions.

  18. Economic benefits of advanced materials in nuclear power systems

    International Nuclear Information System (INIS)

    Busby, J.T.

    2009-01-01

    A key obstacle to the commercial deployment of advanced fast reactors is the capital cost. There is a perception of higher capital cost for fast reactor systems than advanced light water reactors. However, cost estimates come with a large uncertainty since far fewer fast reactors have been built than light water reactor facilities. Furthermore, the large variability of industrial cost estimates complicates accurate comparisons. Reductions in capital cost can result from design simplifications, new technologies that allow reduced capital costs, and simulation techniques that help optimize system design. It is plausible that improved materials will provide opportunities for both simplified design and reduced capital cost. Advanced materials may also allow improved safety and longer component lifetimes. This work examines the potential impact of advanced materials on the capital investment cost of fast nuclear reactors.

  19. Cost-estimate guidelines for advanced nuclear power technologies

    International Nuclear Information System (INIS)

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

    1993-01-01

    Various advanced power plant concepts are currently under development. These include several advanced light water reactors as well as the modular high-temperature gas-cooled reactor and the advanced liquid-metal reactor. One measure-of the attractiveness of a new concept is cost. Invariably, the cost of a new type of power plant will be compared with other alternative forms of electric generation. In order to make reasonable comparative assessments of competing technologies, consistent ground rules and assumptions must be applied when developing cost estimates. This paper describes the cost-estimate guidelines developed by Oak Ridge National Laboratory for the U.S. Department of Energy (DOE) to be used in developing cost estimates for the advanced nuclear reactors and how these guidelines relate to the DOE cost verification process

  20. Online ICPMS detection of the thermal release of fission products from nuclear fuel samples

    International Nuclear Information System (INIS)

    Guenther-Leopold, I.; Svedkauskaite-Le Gore, J.; Kivel, N.

    2009-01-01

    Full text: The release of volatile and semi-volatile fission products (like Cs, Tc, Mo etc.) from spent nuclear fuel by thermal and thermochemical treatment (oxidative or reductive conditions) as a head-end step for advanced reprocessing scenarios is studied in the Hot Laboratory of the Paul Scherrer Institut. For this purpose, a heated sampling cell online connected to an ICPMS (Element 2, Thermo Fisher Scientific) was designed and tested on simulated fuel samples up to 650 o C. The results of this study as well as technical perspectives for heating experiments up to 2000 o C will be presented. (author)

  1. Indian Nuclear Society annual conference-1994 on advanced technologies related to nuclear power: proceedings

    International Nuclear Information System (INIS)

    Grover, R.B.

    1994-01-01

    The focal theme of the conference is advanced technologies related to nuclear power. Over the past three decades civilian nuclear power plants around the world have accumulated about 6000 reactor years of experience and have performed quite well. Overall safety record has been satisfactory. However, nuclear community is trying to compete with its own record by trying to enhance the safety characteristics of the best operating plant. A safety culture has been established in the nuclear establishments, which is providing impetus to advances in all aspects of nuclear technology all over the world. India has ongoing programmes for the development of advanced reactors and related advanced technologies. Evolution of pressurised heavy water reactors in India, developments made in the design of advanced heavy water reactor and the fast reactor programme, are some of the topics covered in addition to highlighting worldwide developments for the next generation of light water reactors. India is one of the few countries in the world where expertise about complete fuel cycle is available. Developments in the back end of the fuel cycle, use of thorium and plutonium and other related issues are also discussed. Technology control regimes being advocated and adopted by developed nations make it imperative for us to indigenise every equipment and component that goes into a power plant. In view of this, some aspects of manufacturing technologies, inspection techniques and maintenance problems are also covered. Relevant papers are processed separately for INIS. (M.K.V.)

  2. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    Stucker, David L.; Miller, Richard S.; Arnsberger, Peter L.

    2004-01-01

    The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

  3. Development of design and analysis software for advanced nuclear system

    International Nuclear Information System (INIS)

    Wu Yican; Hu Liqin; Long Pengcheng; Luo Yuetong; Li Yazhou; Zeng Qin; Lu Lei; Zhang Junjun; Zou Jun; Xu Dezheng; Bai Yunqing; Zhou Tao; Chen Hongli; Peng Lei; Song Yong; Huang Qunying

    2010-01-01

    A series of professional codes, which are necessary software tools and data libraries for advanced nuclear system design and analysis, were developed by the FDS Team, including the codes of automatic modeling, physics and engineering calculation, virtual simulation and visualization, system engineering and safety analysis and the related database management etc. The development of these software series was proposed as an exercise of development of nuclear informatics. This paper introduced the main functions and key techniques of the software series, as well as some tests and practical applications. (authors)

  4. LOFT advanced densitometer for nuclear loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

    1979-01-01

    A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

  5. Advanced Ceramic Materials For Next-Generation Nuclear Applications

    International Nuclear Information System (INIS)

    Marra, J.

    2010-01-01

    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.

  6. 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

  7. COMMIX analysis of four constant flow thermal upramp experiments performed in a thermal hydraulic model of an advanced LMR

    International Nuclear Information System (INIS)

    Yarlagadda, B.S.

    1989-04-01

    The three-dimensional thermal hydraulics computer code COMMIX-1AR was used to analyze four constant flow thermal upramp experiments performed in the thermal hydraulic model of an advanced LMR. An objective of these analyses was the validation of COMMIX-1AR for buoyancy affected flows. The COMMIX calculated temperature histories of some thermocouples in the model were compared with the corresponding measured data. The conclusions of this work are presented. 3 refs., 5 figs

  8. A study on the thermal design of the advanced multi purpose canister

    International Nuclear Information System (INIS)

    Jun, Eun Ju; Chang, S. H.

    2004-01-01

    The storage, transportation and disposal of pressurized water reactor (PWR) spent fuel is an important issue in nuclear industry for safe and extended nuclear power plant operation. However, since wet-storage facilities almost reach to their capacity, dry-storage method, especially using a Multi Purpose Canister (MPC), has been highlighted because it gives better economic and safety point of view. In this study, an advanced design of MPC was proposed through an increase in storage capacity from 26 to 37 PWR fuel assemblies per MPC. Thermal analysis was also carried out by FLUENT 6 code to confirm the temperature criteria of the MPC. From the results, conceptual design of 37 PWR fuel assemblies can reduce the mass per fuel assembly ratio about 19 % compared to the 26 assemblies, and maximum temperature developed at the center of fuel assemblies shows within the range of design criteria. Based on the analysis, it is considered that the advanced MPC design well conforms to the safety requirement and it gives rise to economical gain because of reduction in mass per fuel assembly ratio

  9. Advanced control room design for nuclear power plants

    International Nuclear Information System (INIS)

    Scarola, K.

    1987-01-01

    The power industry has seen a continuous growth of size and complexity of nuclear power plants. Accompanying these changes have been extensive regulatory requirements resulting in significant construction, operation and maintenance costs. In response to related concerns raised by industry members, Combustion Engineering developed the NUPLEX 80 Advanced Control Room. The goal of NUPLEX 80 TM is to: reduce design and construction costs; increase plant safety and availability through improvements in the man-machine interface; and reduce maintenance costs. This paper provides an overview of the NUPLEX 80 Advanced Control Room and explains how the stated goals are achieved. (author)

  10. Advanced robot vision system for nuclear power plants

    International Nuclear Information System (INIS)

    Onoguchi, Kazunori; Kawamura, Atsuro; Nakayama, Ryoichi.

    1991-01-01

    We have developed a robot vision system for advanced robots used in nuclear power plants, under a contract with the Agency of Industrial Science and Technology of the Ministry of International Trade and Industry. This work is part of the large-scale 'advanced robot technology' project. The robot vision system consists of self-location measurement, obstacle detection, and object recognition subsystems, which are activated by a total control subsystem. This paper presents details of these subsystems and the experimental results obtained. (author)

  11. Thermal barrier and support for nuclear reactor fuel core

    International Nuclear Information System (INIS)

    Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

    1987-01-01

    A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

  12. Thermal-hydraulic studies of the Advanced Neutron Source cold source

    International Nuclear Information System (INIS)

    Williams, P.T.; Lucas, A.T.

    1995-08-01

    The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

  13. Nuclear thermal propulsion transportation systems for lunar/Mars exploration

    International Nuclear Information System (INIS)

    Clark, J.S.; Borowski, S.K.; Mcilwain, M.C.; Pellaccio, D.G.

    1992-09-01

    Nuclear thermal propulsion technology development is underway at NASA and DoE for Space Exploration Initiative (SEI) missions to Mars, with initial near-earth flights to validate flight readiness. Several reactor concepts are being considered for these missions, and important selection criteria will be evaluated before final selection of a system. These criteria include: safety and reliability, technical risk, cost, and performance, in that order. Of the concepts evaluated to date, the Nuclear Engine for Rocket Vehicle Applications (NERVA) derivative (NDR) is the only concept that has demonstrated full power, life, and performance in actual reactor tests. Other concepts will require significant design work and must demonstrate proof-of-concept. Technical risk, and hence, development cost should therefore be lowest for the concept, and the NDR concept is currently being considered for the initial SEI missions. As lighter weight, higher performance systems are developed and validated, including appropriate safety and astronaut-rating requirements, they will be considered to support future SEI application. A space transportation system using a modular nuclear thermal rocket (NTR) system for lunar and Mars missions is expected to result in significant life cycle cost savings. Finally, several key issues remain for NTR's, including public acceptance and operational issues. Nonetheless, NTR's are believed to be the next generation of space propulsion systems - the key to space exploration

  14. Objectives for the development of advanced nuclear plants

    International Nuclear Information System (INIS)

    1993-01-01

    The scope of this report is to reiterate the broad objectives for the development of advanced nuclear plants, to set forth some related subobjectives and to propose some universal goals for the development programmes. The majority of these can be categorized under the headings of enhancing safety, improving reliability and gaining better economics. These categories are used in the report followed by additional categories considered to be important within the global framework intended. Additional broad objectives appear unlikely but more subobjectives may become evident as time progresses and the need arises to express them in the intended more global framework. The goals also may change. The scope is therefore a set of objectives for development of advanced nuclear plants. The objectives are believed to be universally acceptable; they have been reviewed on that basis. 13 refs

  15. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  16. A computational model for thermal fluid design analysis of nuclear thermal rockets

    International Nuclear Information System (INIS)

    Given, J.A.; Anghaie, S.

    1997-01-01

    A computational model for simulation and design analysis of nuclear thermal propulsion systems has been developed. The model simulates a full-topping expander cycle engine system and the thermofluid dynamics of the core coolant flow, accounting for the real gas properties of the hydrogen propellant/coolant throughout the system. Core thermofluid studies reveal that near-wall heat transfer models currently available may not be applicable to conditions encountered within some nuclear rocket cores. Additionally, the possibility of a core thermal fluid instability at low mass fluxes and the effects of the core power distribution are investigated. Results indicate that for tubular core coolant channels, thermal fluid instability is not an issue within the possible range of operating conditions in these systems. Findings also show the advantages of having a nonflat centrally peaking axial core power profile from a fluid dynamic standpoint. The effects of rocket operating conditions on system performance are also investigated. Results show that high temperature and low pressure operation is limited by core structural considerations, while low temperature and high pressure operation is limited by system performance constraints. The utility of these programs for finding these operational limits, optimum operating conditions, and thermal fluid effects is demonstrated

  17. 15 N utilization in nitride nuclear fuels for advanced nuclear power reactors and accelerator - driven systems

    International Nuclear Information System (INIS)

    Axente, D.

    2005-01-01

    15 N utilization for nitride nuclear fuels production for nuclear power reactors and accelerator - driven systems is presented. Nitride nuclear fuel is the obvious choice for advanced nuclear reactors and ADS because of its favorable properties: a high melting point, excellent thermal conductivity, high fissile density, lower fission gas release and good radiation tolerance. The application of nitride fuels in nuclear reactors and ADS requires use of 15 N enriched nitrogen to suppress 14 C production due to (n,p) reaction on 14 N. Accelerator - driven system is a recent development merging of accelerator and fission reactor technologies to generate electricity and transmute long - lived radioactive wastes as minor actinides: Np, Am, Cm. A high-energy proton beam hitting a heavy metal target produces neutrons by spallation. The neutrons cause fission in the fuel, but unlike in conventional reactors, the fuel is sub-critical and fission ceases when the accelerator is turned off. Nitride fuel is a promising candidate for transmutation in ADS of minor actinides, which are converted into nitrides with 15 N for that purpose. Tacking into account that the world wide market is about 20 to 40 Kg 15 N annually, the supply of that isotope for nitride fuel production for nuclear power reactors and ADS would therefore demand an increase in production capacity by a factor of 1000. For an industrial plant producing 100 t/y 15 N, using present technology of isotopic exchange in NITROX system, the first separation stage of the cascade would be fed with 10M HNO 3 solution of 600 mc/h flow - rate. If conversion of HNO 3 into NO, NO 2 , at the enriching end of the columns, would be done with gaseous SO 2 , for a production plant of 100 t/y 15 N a consumption of 4 million t SO 2 /y and a production of 70 % H 2 SO 4 waste solution of 4.5 million mc/y are estimated. The reconversion of H 2 SO 4 into SO 2 in order to recycle of SO 2 is a problem to be solved to compensate the cost of SO 2

  18. Development of advanced materials and devices for nuclear radiation measurements

    International Nuclear Information System (INIS)

    Gadkari, S.C.

    2015-01-01

    Single crystals of technologically important materials are grown in the Crystal Technology Section of the Technical Physics Division, BARC. These crystals find applications as scintillators and dosimeters in nuclear radiation detection/measurements. Scintillator crystals of some advanced materials like cerium doped Gd 3 Ga 3 Al 2 O 12 , Lu 2 SiO 5 , YAIO 3 etc and some conventional materials such as Bi 4 Ge 3 O 12 , CsI:Tl, NaI:Tl, etc have been grown from melts using the Czochralski and Bridgman techniques. Portable gamma-ray spectrometers that work from a USB port of a laptop have been developed using the grown scintillator crystals. In recent years there has been a flurry of research activities on materials containing Li 6 , B 10 , etc that have large capture cross-sections for neutrons to develop solid state detectors for neutrons. For this purpose single crystals of cerium doped Li 6 Y(BO 3 ) 3 and silver doped Li 2 B 4 O 7 have been developed. Optical, thermo-luminescence, photo-luminescence and scintillation properties of these crystals have been investigated with a view to develop detectors and dosimeters. The Li 2 B 4 O 7 :Ag is a tissue equivalent material (Z eff = 7.3 close to 7.4 of tissue) useful in the personal and medical dosimetry applications. As the emission of Ag + lies in the UV region (267 nm), a customized TL measurement set-up has been developed using a solar blind PMT that enabled the measurement of very low doses below 5 μGy and linearity up to 100 Gy. Films of CsI:TI in the 10 nm to 3 μm thickness range were deposited on silicon substrates using the physical vapor deposition technique under vacuum conditions. The deposited films investigated using SEM and AFM revealed a columnar growth behavior with a preferential orientation along <200>. The growth of single crystals from melts, recent efforts in the development of detectors and results of experiments conducted to detect thermal neutrons are described. (author)

  19. Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

    Science.gov (United States)

    Emrich, William J., Jr.; Pearson, J. Boise; Schoenfeld, Michael P.

    2015-01-01

    The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at

  20. 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

  1. Advanced nuclear plant design options to cope with external events

    International Nuclear Information System (INIS)

    2006-02-01

    With the stagnation period of nuclear power apparently coming to an end, there is a renewed interest in many Member States in the development and application of nuclear power plants (NPPs) with advanced reactors. Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR Finland for Finland and France) and more are under consideration. There is a noticeable progress in the development and demonstration of innovative high temperature gas cooled reactors, for example, in China, South Africa and Japan. The Generation IV International Forum has defined the International Near Term Deployment programme and, for a more distant perspective, six innovative nuclear energy systems have been selected and certain R and D started by several participating countries. National efforts on design and technology development for NPPs with advanced reactors, both evolutionary and innovative, are ongoing in many Member States. Advanced NPPs have an opportunity to be built at many sites around the world, with very broad siting conditions. There are special concerns that safety of these advanced reactors may be challenged by external events following new scenarios and failure modes, different from those well known for the currently operated reactors. Therefore, the engineering community identified the need to assess the proposed design configurations in relation to external scenarios at the earliest stages of the design development. It appears that an early design optimization in relation to external events is a necessary requirement to achieve safe and economical advanced nuclear power plants. Reflecting on these developments, the IAEA has planned the preparation of a report to define design options for protection from external event impacts in NPPs with evolutionary and innovative reactors. The objective of this publication is to present the state-of-the-art in design approaches for the protection of NPPs with evolutionary and innovative

  2. Advanced ceramic materials for next-generation nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Marra, John [Savannah River National Laboratory Aiken, SC 29802 (United States)

    2011-10-29

    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

  3. Technical verification of advanced nuclear fuel for KSNPs

    International Nuclear Information System (INIS)

    Lee, C. B.; Bang, J. G.; Kim, D. H. and others

    2002-03-01

    KNFC has developed the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants through the three-year R and D project (from April 1999 to March 2002) under the Nuclear R and D program by MOST. The purpose of this project is to verify the advanced 16x16 fuel assembly for the Korean Standard Nuclear Plants being developed by KNFC during the same period. Verification tests for the advanced fuel assembly and its components such as characteristic test on the spacer grid spring and dimple, static buckling and dynamic impact test on the 5x5 partial spacer grid, the fuel rod vibration test supported by the PLUS7 mid-spacer grid, fretting wear test, turbulent flow structure test in wind tunnel and corrosion test were performed by using the KAERI facilities. Design reports and test results produced by KNFC were technically reviewed. For the domestic production of burnable poison rod, manufacturing technology of burnable poison pellets was developed

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

    International Nuclear Information System (INIS)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-02-01

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

  6. Need for nuclear data for thermal neutron reactors

    International Nuclear Information System (INIS)

    Bouchard, J.; Golinelli, C.; Tellier, H.

    1983-01-01

    The need for nuclear data for thermal neutron reactors is conditioned by the persisting lack of agreement between the calculation and measurement of certain parameters, by the benefit that can be drawn from reduction of the marginal areas and by envisaged modifications. Three particular fields are delineated. Reduction of the deviation in temperature coefficients by modification of the shape of the effective capture cross sections of uranium-238 and -235 in the thermal range. The increase in precision of kinetic measurements by a better knowledge of data connected to slowed-down neutrons. Improvement in predicting the neutron activity of the fuels used in measuring the effective capture cross sections of plutonium-242 and americium-243. (Auth.)

  7. 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.

  8. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  9. Terms for describing new, advanced nuclear power plants

    International Nuclear Information System (INIS)

    1997-04-01

    The IAEA's Division of Nuclear Power and the Fuel Cycle (then the Division of Nuclear Power) took an initiative in this field some years ago when work was initiated in the area of ''safety related terms'' by its International Working Group on Advanced Technologies for Water Cooled Reactors. This activity drew on advice from reactor design organizations, research institutes and government organizations, and aimed at helping eliminate confusion and misuse of safety related terms in widespread use, clarifying technical thinking regarding these terms, and improving nuclear power acceptability by providing precisely described technical meanings to them. After discussion also in the International Working Groups for Gas Cooled Reactors and Fast Reactors, the work resulted in the publication in September 1991 of IAEA-TECDOC-626, entitled ''Safety Related Terms for Advanced Nuclear Plants'', which has become a widely used publication. The present TECDOC has been prepared using the same approach to obtain advice from involved parties. Drafts of this report have been reviewed by the International Working Groups on Water Cooled Reactors, Fast Reactors and Gas Cooled Reactors, as well as by the IAEA's International Fusion Research Council (IFRC). The comments and suggestions received have been evaluated and utilized for producing the present TECDOC. 3 figs

  10. 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.

  11. Analyses of thermal plume of Cernavoda nuclear power plant by satellite remote sensing data

    Science.gov (United States)

    Zoran, M. A.; Nicolae, D. N.; Talianu, C. L.; Ciobanu, M.; Ciuciu, J. G.

    2005-10-01

    The synergistic use of multi-temporal and multi-spectral remote sensing data offers the possibility of monitoring of environment quality in the vicinity of nuclear power plants (NPP). Advanced digital processing techniques applied to several LANDSAT, MODIS and ASTER data are used to assess the extent and magnitude of radiation and non-radiation effects on the water, near field soil, vegetation and air for NPP Cernavoda , Romania . Cernavoda Unit 1 power plant, using CANDU technology, having 706.5 MW power, is successfully in operation since 1996. Cernavoda Unit 2 which is currently under construction will be operational in 2007. Thermal discharge from nuclear reactor cooling is dissipated as waste heat in Danube-Black -Sea Canal and Danube river. Water temperature distributions captured in thermal IR imagery are correlated with meteorological parameters. Additional information regarding flooding events and earthquake risks is considered . During the winter, the thermal plume is localized to an area within a few km of the power plant, and the temperature difference between the plume and non-plume areas is about 1.5 oC. During the summer and fall, there is a larger thermal plume extending 5-6 km far along Danube Black Sea Canal, and the temperature change is about 1.0 oC. Variation of surface water temperature in the thermal plume is analyzed. The strong seasonal difference in the thermal plume is related to vertical mixing of the water column in winter and to stratification in summer. Hydrodynamic simulation leads to better understanding of the mechanisms by which waste heat from NPP Cernavoda is dissipated in the environment.

  12. Shear viscosity and thermal conductivity of nuclear 'pasta'

    International Nuclear Information System (INIS)

    Horowitz, C. J.; Berry, D. K.

    2008-01-01

    We calculate the shear viscosity η and thermal conductivity κ of a nuclear pasta phase in neutron star crusts. This involves complex nonspherical shapes. We use semiclassical molecular dynamics simulations involving 40, 000 to 100, 000 nucleons. The viscosity η can be simply expressed in terms of the height Z* and width Δq of the peak in the static structure factor S p (q). We find that η increases somewhat, compared to a lower density phase involving spherical nuclei, because Z* decreases from form factor and ion screening effects. However, we do not find a dramatic increase in η from nonspherical shapes, as may occur in conventional complex fluids

  13. Smart built-in test for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Lombrozo, P.C.

    1992-03-01

    Smart built-in test (BIT) technologies are envisioned for nuclear thermal propulsion spacecraft components which undergo constant irradiation and are therefore unsafe for manual testing. Smart BIT systems of automated/remote type allow component and system tests to be conducted; failure detections are directly followed by reconfiguration of the components affected. The 'smartness' of the BIT system in question involves the reduction of sensor counts via the use of multifunction sensors, the use of components as integral sensors, and the use of system design techniques which allow the verification of system function beyond component connectivity

  14. New water intake systems for thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Ishchuk, T.B.; Samodel'nikov, B.T.

    1989-01-01

    Problems arising during design of water intake and spillway structures for the auxiliary water supply system of thermal and nuclear power plants connected with the provision of their reliable operation and with the effect on the temperature condition of reservoirs and their ecology are investigated. Design providing for the connection of intake channel and catch drain for a through (transition) channel and supplying a water transition flow by ejecting water outputs is suggested. The variant considered is effective for seas, lakes and reservoirs with adverse conditions for natural cooling and it is suitable for regions with seismicity up to 5-6 balls

  15. A cermet fuel reactor for nuclear thermal propulsion

    Science.gov (United States)

    Kruger, Gordon

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

  16. A cermet fuel reactor for nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Kruger, G.

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk

  17. Thermal radiation in gas core nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J.

    1994-01-01

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs

  18. Korean views on needs for international cooperation in development and development of advanced nuclear power systems

    International Nuclear Information System (INIS)

    Yoon, Young Ku; Lee, Byong Whi; Shim, Chang Saeng.

    1993-01-01

    Korea methodology and experience in international cooperation in the field of construction and operation of nuclear power plants as well as Korean views on development and deployment of advanced nuclear nuclear power systems are presented

  19. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

    Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.

  20. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

    Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers deal with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately

  1. Advanced nuclear fuel cycles and radioactive waste management

    International Nuclear Information System (INIS)

    2006-01-01

    This study analyses a range of advanced nuclear fuel cycle options from the perspective of their effect on radioactive waste management policies. It presents various fuel cycle options which illustrate differences between alternative technologies, but does not purport to cover all foreseeable future fuel cycles. The analysis extends the work carried out in previous studies, assesses the fuel cycles as a whole, including all radioactive waste generated at each step of the cycles, and covers high-level waste repository performance for the different fuel cycles considered. The estimates of quantities and types of waste arising from advanced fuel cycles are based on best available data and experts' judgement. The effects of various advanced fuel cycles on the management of radioactive waste are assessed relative to current technologies and options, using tools such as repository performance analysis and cost studies. (author)

  2. Scaling in nuclear reactor system thermal-hydraulics

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.M.

    2010-01-01

    Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.

  3. Scaling in nuclear reactor system thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    D' Auria, F., E-mail: dauria@ing.unipi.i [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Galassi, G.M. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, University of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)

    2010-10-15

    Scaling is a reference 'key-word' in engineering and in physics. The relevance of scaling in the water cooled nuclear reactor technology constitutes the motivation for the present paper. The origin of the scaling-issue, i.e. the impossibility to get access to measured data in case of accident in nuclear reactors, is discussed at first. The so-called 'scaling-controversy' constitutes an outcome. Then, a critical survey (or 'scaling state-of-art';) is given of the attempts and of the approaches to provide a solution to the scaling-issue in the area of Nuclear Reactor System Thermal-Hydraulics (NRSTH): dimensionless design factors for Integral Test Facilities (ITF) are distinguished from scaling factors. The last part of the paper has a two-fold nature: (a) classifying the information about achievements in the area of thermal-hydraulics which are relevant to scaling: the concepts of 'scaling-pyramid' and the related 'scaling bridges' are introduced; (b) establishing a logical path across the scaling achievements (represented as a 'scaling puzzle'). In this context, the 'roadmap for scaling' is proposed: the objective is addressing the scaling issue when demonstrating the applicability of system codes in the licensing process of nuclear power plants. The code itself is referred hereafter as the 'key-to-scaling'. The database from the operation of properly scaled ITF and the availability of qualified system codes are identified as main achievements in NRSTH connected with scaling. The 'roadmap to scaling' constitutes a unified approach to scaling which aims at solving the 'scaling puzzle' created by researches performed during a half-a-century period.

  4. Advanced nuclear fuel cycles - Main challenges and strategic choices

    International Nuclear Information System (INIS)

    Le Biez, V.; Machiels, A.; Sowder, A.

    2013-01-01

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness

  5. Advanced nuclear fuel cycles - Main challenges and strategic choices

    Energy Technology Data Exchange (ETDEWEB)

    Le Biez, V. [Corps des Mines, 35 bis rue Saint-Sabin, F-75011 Paris (France); Machiels, A.; Sowder, A. [Electric Power Research Institute, Inc. 3420, Hillview Avenue, Palo Alto, CA 94304 (United States)

    2013-07-01

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness.

  6. Thermal-hydraulic software development for nuclear waste transportation cask design and analysis

    International Nuclear Information System (INIS)

    Brown, N.N.; Burns, S.P.; Gianoulakis, S.E.; Klein, D.E.

    1991-01-01

    This paper describes the development of a state-of-the-art thermal-hydraulic software package intended for spent fuel and high-level nuclear waste transportation cask design and analysis. The objectives of this software development effort are threefold: (1) to take advantage of advancements in computer hardware and software to provide a more efficient user interface, (2) to provide a tool for reducing inefficient conservatism in spent fuel and high-level waste shipping cask design by including convection as well as conduction and radiation heat transfer modeling capabilities, and (3) to provide a thermal-hydraulic analysis package which is developed under a rigorous quality assurance program established at Sandia National Laboratories. 20 refs., 5 figs., 2 tabs

  7. Transient thermal creep of nuclear reactor pressure vessel type concretes

    International Nuclear Information System (INIS)

    Khoury, G.A.

    1983-01-01

    The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and

  8. Investigation of transient thermal dissipation in thinned LSI for advanced packaging

    Science.gov (United States)

    Araga, Yuuki; Shimamoto, Haruo; Melamed, Samson; Kikuchi, Katsuya; Aoyagi, Masahiro

    2018-04-01

    Thinning of LSI is necessary for superior form factor and performance in dense cutting-edge packaging technologies. At the same time, degradation of thermal characteristics caused by the steep thermal gradient on LSIs with thinned base silicon is a concern. To manage a thermal environment in advanced packages, thermal characteristics of the thinned LSIs must be clarified. In this study, static and dynamic thermal dissipations were analyzed before and after thinning silicon to determine variations of thermal characteristics in thinned LSI. Measurement results revealed that silicon thinning affects dynamic thermal characteristics as well as static one. The transient variations of thermal characteristics of thinned LSI are precisely verified by analysis using an equivalent model based on the thermal network method. The results of analysis suggest that transient thermal characteristics can be easily estimated by employing the equivalent model.

  9. Thermal conductivity thermal diffusivity of UO{sub 2}-BeO nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Mansur, Fábio A.; Camarano, Denise M.; Santos, Ana M. M.; Ferraz, Wilmar B.; Silva, Mayra A.; Ferreira, Ricardo A.N., E-mail: fam@cdtn.br, E-mail: dmc@cdtn.br, E-mail: amms@cdtn.br, E-mail: ferrazw@cdtn.br, E-mail: mayra.silva@cdtn.br, E-mail: ricardoanf@yahoo.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The temperature distribution in nuclear fuel pellets is of vital importance for the performance of the reactor, as it affects the heat transfer, the mechanical behavior and the release of fission gas during irradiation, reducing safety margins in possible accident scenarios. One of the main limitation for the current uranium dioxide nuclear fuel (UO{sub 2}) is its low thermal conductivity, responsible for the higher temperature of the pellet center and, consequently, for a higher radial temperature gradient. Thus, the addition of another material to increase the UO{sub 2} fuel thermal conductivity has been considered. Among the additives that are being investigated, beryllium oxide (BeO) has been chosen due to its high thermal conductivity, with potential to optimize power generation in pressurized light water reactors (PWR). In this work, UO{sub 2}-BeO pellets were obtained by the physical mixing of the powders with additions of 2wt% and 3wt% of BeO. The thermal diffusivity and conductivity of the pellets were determined from room temperature up to 500 °C. The results were normalized to 95% of the theoretical density (TD) of the pellets and varied according to the BeO content. The range of the values of thermal diffusivity and conductivity were 1.22 mm{sup 2}∙s{sup -1} to 3.69 mm{sup 2}∙s{sup -1} and 3.80 W∙m{sup -}'1∙K{sup -1} to 9.36 W∙m{sup -1}∙K{sup -1}, respectively. (author)

  10. Research and development of advanced robots for nuclear power plants

    International Nuclear Information System (INIS)

    Tsukune, Hideo; Hirukawa, Hirohisa; Kitagaki, Kosei; Liu, Yunhui; Onda, Hiromu; Nakamura, Akira

    1994-01-01

    Social and economic demands have been pressing for automation of inspection tasks, maintenance and repair jobs of nuclear power plants, which are carried out by human workers under circumstances with high radiation level. Since the plants are not always designed for introduction of automatic machinery, sophisticated robots shall play a crucial role to free workers from hostile environments. We have been studying intelligent robot systems and regarded nuclear industries as one of the important application fields where we can validate the feasibility of the methods and systems we have developed. In this paper we firstly discuss on the tasks required in nuclear power plants. Secondly we introduce current status of R and D on special purpose robots, versatile robots and intelligent robots for automatizing the tasks. Then we focus our discussions on three major functions in realizing robotized assembly tasks under such unstructured environments as in nuclear power plants; planning, vision and manipulation. Finally we depict an image of a prototype robot system for nuclear power plants based on the advanced functions. (author) 64 refs

  11. Characterisation of advanced windows. Determination of thermal properties by measurements

    Energy Technology Data Exchange (ETDEWEB)

    Duer, K.

    2001-04-01

    This report describes work carried out with the aim of facilitating a full energy performance characterisation of advanced windows and glazings by means of measurements. The energy performance of windows and glazings are characterised by two parameters: The thermal transmittance (U-value) and the total solar energy transmittance (g-value) and methods to determine these two parameters by measurements have been investigated. This process has included the improvement of existing equipment and existing measuring methods as well as the development of new measuring equipment and new methods of measuring and data treatment. Measurements of the thermal transmittance of windows and glazings in a guarded hot box have been investigated. The calibration and measuring procedures for determining the U-values of facade windows were analysed and a suggestion for a new calibration and measuring procedure for determining the U-values of roof windows in a guarded hot box was elaborated. The accuracy of the guarded hot box measurements was examined by comparisons to measurements in a hot-plate device and excellent agreement between the results was obtained. Analysis showed that the expected uncertainty in the U-value measurement is about 5% for a specimen with a U-value of 1.75 W/m{sup 2}K. The U-values of three different windows were measured in two separate round robin tests applying two different calibration procedures. The windows U-values where ranging from 1.1 to 2.5 W/m{sup 2}K and all measured results were within the expected uncertainties of the measurements. On the basis of the investigations on hot box measurements a high degree of confidence in the measurement accuracy and the measuring procedure of the guarded hot box at the Department of Buildings and Energy has been obtained. Indoor g-value measurements in a calorimetric test facility (the METSET) mounted in a solar simulator have been investigated and a number of problems regarding these measurements have been

  12. Benchmarking of thermalhydraulic loop models for lead-alloy-cooled advanced nuclear energy systems. Phase I: Isothermal forced convection case

    International Nuclear Information System (INIS)

    2012-06-01

    Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of the Fuel Cycle (WPFC) has been established to co-ordinate scientific activities regarding various existing and advanced nuclear fuel cycles, including advanced reactor systems, associated chemistry and flowsheets, development and performance of fuel and materials and accelerators and spallation targets. The WPFC has different expert groups to cover a wide range of scientific issues in the field of nuclear fuel cycle. The Task Force on Lead-Alloy-Cooled Advanced Nuclear Energy Systems (LACANES) was created in 2006 to study thermal-hydraulic characteristics of heavy liquid metal coolant loop. The objectives of the task force are to (1) validate thermal-hydraulic loop models for application to LACANES design analysis in participating organisations, by benchmarking with a set of well-characterised lead-alloy coolant loop test data, (2) establish guidelines for quantifying thermal-hydraulic modelling parameters related to friction and heat transfer by lead-alloy coolant and (3) identify specific issues, either in modelling and/or in loop testing, which need to be addressed via possible future work. Nine participants from seven different institutes participated in the first phase of the benchmark. This report provides details of the benchmark specifications, method and code characteristics and results of the preliminary study: pressure loss coefficient and Phase-I. A comparison and analysis of the results will be performed together with Phase-II

  13. Proceedings of the 2008 International Congress on Advances in Nuclear Power Plants - ICAPP '08

    International Nuclear Information System (INIS)

    2008-01-01

    ICAPP 2008 congress brought together international experts of the nuclear industry involved in the operation, development, building, regulation and research related to Nuclear Power Plants. The program covered the full spectrum of Nuclear Power Plant issues from design, deployment and construction of plants to research and development of future designs and advanced systems. It covered also lessons learned from power, research and demonstration reactors from over 50 years of experience with operation and maintenance, structures, materials, technical specifications, human factors, system design and reliability. The program comprised 13 technical tracks: 1. Water-Cooled Reactor Programs and Issues: Evolutionary designs, innovative, passive, light and heavy water cooled reactors; issues related to meeting near term utility needs; design issues; business, economical cost challenges; infrastructure limitations and improved construction techniques including modularization. 2. High Temperature Gas Cooled Reactors: Design and development issues, components and materials, safety, reliability, economics, demonstration plants and environmental issues, fuel design and reliability, power conversion technology, impact of non electricity applications on reactor design; advanced thermal and fast reactors. 3. LMFR and Longer Term Reactor Programs: Reactor technology with enhanced fuel cycle features for improved resource utilization, waste characteristics, and power conversion capabilities. Potential reactor designs with longer development times such as super critical water reactors and liquid fuel reactors, Gen IV, INPRO, EUR and other programs. 4. Operation, Performance and Reliability Management: Training, O and M costs, life cycle management, risk based maintenance, operational experiences, performance and reliability improvements, outage optimization, human factors, plant staffing, outage reduction features, major component reliability, repair and replacement, in

  14. Turbopump Design and Analysis Approach for Nuclear Thermal Rockets

    International Nuclear Information System (INIS)

    Chen, Shucheng S.; Veres, Joseph P.; Fittje, James E.

    2006-01-01

    A rocket propulsion system, whether it is a chemical rocket or a nuclear thermal rocket, is fairly complex in detail but rather simple in principle. Among all the interacting parts, three components stand out: they are pumps and turbines (turbopumps), and the thrust chamber. To obtain an understanding of the overall rocket propulsion system characteristics, one starts from analyzing the interactions among these three components. It is therefore of utmost importance to be able to satisfactorily characterize the turbopump, level by level, at all phases of a vehicle design cycle. Here at the NASA Glenn Research Center, as the starting phase of a rocket engine design, specifically a Nuclear Thermal Rocket Engine design, we adopted the approach of using a high level system cycle analysis code (NESS) to obtain an initial analysis of the operational characteristics of a turbopump required in the propulsion system. A set of turbopump design codes (PumpDes and TurbDes) were then executed to obtain sizing and performance parameters of the turbopump that were consistent with the mission requirements. A set of turbopump analyses codes (PUMPA and TURBA) were applied to obtain the full performance map for each of the turbopump components; a two dimensional layout of the turbopump based on these mean line analyses was also generated. Adequacy of the turbopump conceptual design will later be determined by further analyses and evaluation. In this paper, descriptions and discussions of the aforementioned approach are provided and future outlooks are discussed

  15. Advanced Nuclear Applications in Medicinr at Chiang Mai University

    International Nuclear Information System (INIS)

    Vilasdechanon, Nonglak

    2015-01-01

    The atomic energy applications in Faculty of Medicine, Chiang Mai University (CMU) are mainly performed by department of Radiology that is divided into three dicisions: 1) Diagnostic Radiology Division for the applications of X-rays, ultrasound, and magnetic resonance, 2) Therapeutic Radiology and Oncology Division for cancer treatments by photon accelrator and external radionuclides therapy or brachytherapy, 3) Nuclear Meddicine Division for clinical dignosis by using radionuclide scintigraphy, targeted molecular imaging and internal radionuclide therapy. In the last decade, many advanced medical images for clinical diagnosis included of digital & computed radiology (DR & CR), digital subtraction angiography (DSA) and images (DSI), computed tomography (CT) with dual X-rays energies, manetic resonance imaging (MRI), and hybrid images of SPECT/CT were established in Radiology Department and PET/CT Cyclitron Center Chiang Mai University (PCCMU), respectively. For cancer treatments, the frontier technologies in radiation oncligy therapy such as tomotherapy, IMRT, 3D conformal radiation treatment, stereotactic radiationtherapy (SRT), stereotactic radiation surgery (SRS), and radiation biology laboratory were implemented in the department as well. As far as fast development of nuclear technology in medicine, future implementation of advanced nuclear applications in medicine strongly need an intergrated knoowledge from many specialties e.g. computer softeare in image reconstruction, accuracy and and precision technology, production of specific radiotracers for molecular imaging and therapy, techniques in radionuclide productions, innovation of new wquipment or materials for radiationprotection and safety, etc. However the most important factors of nuclear applications in medicine are the vision, mission and the value statements of the organization on the high cost in radiology investment and human resources development. We have to emphasize that people who are involved

  16. General theory for thermal pulses of finite amplitude in nuclear shell-burnings

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, D [Tokyo Univ. (Japan). Coll. of General Education; Fujimoto, M Y

    1978-09-01

    Theory for thermal pulses of nuclear shell-burning is advanced to include the case of finite amplitude. The aims are to predict the progress of thermal pulse quantitatively and to obtain the peak values of the temperature and nuclear energy generation rate without making detailed numerical computation of stellar structure. In order to attain them the physical processes involved in the progress of the pulse are clarified using the concepts of the flatness of the shell source, which destabilizes nuclear burning, and the effect of radiation pressure, which stabilizes it. It is shown that the progress of the pulse can be predicted quantitatively when the pressure and the gravitational potential of the burning shell are specified for the onset stage of the pulse. The pulse height is determined mainly by the initial pressure; the higher initial pressure results in the higher pulse. Mass dependence is also obtained by approximating the gravitational potential by that of white dwarfs. The initial pressure is the quantity which is determined in the course of evolution preceding the pulse. The theory is shown to give a satisfactory agreement with numerical computations for a wide variety of the preceding evolutions, i.e., both for the case of the core in red giant stars and of the accreting white dwarfs.

  17. The Space Nuclear Thermal Propulsion Program: Propulsion for the twenty first century

    International Nuclear Information System (INIS)

    Bleeker, G.; Moody, J.; Kesaree, M.

    1993-01-01

    As mission requirements approach the limits of the chemical propulsion systems, new engines must be investigated that can meet the advanced mission requirements of higher payload fractions, higher velocities, and consequently higher specific Impulses (Isp). The propulsion system that can meet these high demands is a nuclear thermal rocket engine. This engine generates the thrust by expanding/existing the hydrogen, heated from the energy derived from the fission process in a reactor, through a nozzle. The Department of Defense (DoD), however, initiated a new nuclear rocket development program in 1987 for ballistic missile defense application. The Space Nuclear Thermal Propulsion (SNTP) Program that seeks to improve on the technology of ROVER/NERVA grew out of this beginning and has been managed by the Air Force, with the involvement of DoE and NASA. The goal of the SNTP Program is to develop an engine to meet potential Air Force requirements for upper stage engine, bimodal propulsion/power applications, and orbital transfer vehicles, as well as the NASA requirements for possible missions to the Moon and Mars. During the entire life of the program, the DoD has considered safety to be of paramount importance, and is following all national environmental policies

  18. Numerical analysis and nuclear standard code application to thermal fatigue

    International Nuclear Information System (INIS)

    Merola, M.

    1992-01-01

    The present work describes the Joint Research Centre Ispra contribution to the IAEA benchmark exercise 'Lifetime Behaviour of the First Wall of Fusion Machines'. The results of the numerical analysis of the reference thermal fatigue experiment are presented. Then a discussion on the numerical analysis of thermal stress is tackled, pointing out its particular aspects in view of their influence on the stress field evaluation. As far as the design-allowable number of cycles are concerned the American nuclear code ASME and the French code RCC-MR are applied and the reasons for the different results obtained are investigated. As regards a realistic fatigue lifetime evaluation, the main problems to be solved are brought out. This work, is intended as a preliminary basis for a discussion focusing on the main characteristics of the thermal fatigue problem from both a numerical and a lifetime assessment point of view. In fact the present margin of discretion left to the analyst may cause undue discrepancies in the results obtained. A sensitivity analysis of the main parameters involved is desirable and more precise design procedures should be stated

  19. Testing for Nuclear Thermal Propulsion Systems: Identification of Technologies for Effluent Treatment in Test Facilities

    Data.gov (United States)

    National Aeronautics and Space Administration — Key steps to ensure identification of relevant effluent treatment technologies for Nuclear Thermal Propulsion (NTP) testing include the following. 1. Review of...

  20. Fabrication and Testing of CERMET Fuel Materials for Nuclear Thermal Propulsion

    Science.gov (United States)

    Hickman, Robert; Broadway, Jeramie; Mireles, Omar

    2012-01-01

    A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on Nuclear Thermal Propulsion (NTP) is currently being developed for Advanced Space Exploration Systems. The overall goal of the project is to address critical NTP technology challenges and programmatic issues to establish confidence in the affordability and viability of NTP systems. The current technology roadmap for NTP identifies the development of a robust fuel form as a critical near term need. The lack of a qualified nuclear fuel is a significant technical risk that will require a considerable fraction of program resources to mitigate. Due to these risks and the cost for qualification, the development and selection of a primary fuel must begin prior to Authority to Proceed (ATP) for a specific mission. The fuel development is a progressive approach to incrementally reduce risk, converge the fuel materials, and mature the design and fabrication process of the fuel element. A key objective of the current project is to advance the maturity of CERMET fuels. The work includes fuel processing development and characterization, fuel specimen hot hydrogen screening, and prototypic fuel element testing. Early fuel materials development is critical to help validate requirements and fuel performance. The purpose of this paper is to provide an overview and status of the work at Marshall Space Flight Center (MSFC).

  1. Advanced deposition model for thermal activated chemical vapor deposition

    Science.gov (United States)

    Cai, Dang

    Thermal Activated Chemical Vapor Deposition (TACVD) is defined as the formation of a stable solid product on a heated substrate surface from chemical reactions and/or dissociation of gaseous reactants in an activated environment. It has become an essential process for producing solid film, bulk material, coating, fibers, powders and monolithic components. Global market of CVD products has reached multi billions dollars for each year. In the recent years CVD process has been extensively used to manufacture semiconductors and other electronic components such as polysilicon, AlN and GaN. Extensive research effort has been directed to improve deposition quality and throughput. To obtain fast and high quality deposition, operational conditions such as temperature, pressure, fluid velocity and species concentration and geometry conditions such as source-substrate distance need to be well controlled in a CVD system. This thesis will focus on design of CVD processes through understanding the transport and reaction phenomena in the growth reactor. Since the in situ monitor is almost impossible for CVD reactor, many industrial resources have been expended to determine the optimum design by semi-empirical methods and trial-and-error procedures. This approach has allowed the achievement of improvements in the deposition sequence, but begins to show its limitations, as this method cannot always fulfill the more and more stringent specifications of the industry. To resolve this problem, numerical simulation is widely used in studying the growth techniques. The difficulty of numerical simulation of TACVD crystal growth process lies in the simulation of gas phase and surface reactions, especially the latter one, due to the fact that very limited kinetic information is available in the open literature. In this thesis, an advanced deposition model was developed to study the multi-component fluid flow, homogeneous gas phase reactions inside the reactor chamber, heterogeneous surface

  2. Nuclear desalination in the Arab world - Part II: Advanced inherent and passive safe nuclear reactors

    International Nuclear Information System (INIS)

    Karameldin, A.; Samer S. Mekhemar

    2004-01-01

    Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive. (author)

  3. Uncertainty and sensitivity analysis of the nuclear fuel thermal behavior

    Energy Technology Data Exchange (ETDEWEB)

    Boulore, A., E-mail: antoine.boulore@cea.fr [Commissariat a l' Energie Atomique (CEA), DEN, Fuel Research Department, 13108 Saint-Paul-lez-Durance (France); Struzik, C. [Commissariat a l' Energie Atomique (CEA), DEN, Fuel Research Department, 13108 Saint-Paul-lez-Durance (France); Gaudier, F. [Commissariat a l' Energie Atomique (CEA), DEN, Systems and Structure Modeling Department, 91191 Gif-sur-Yvette (France)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer A complete quantitative method for uncertainty propagation and sensitivity analysis is applied. Black-Right-Pointing-Pointer The thermal conductivity of UO{sub 2} is modeled as a random variable. Black-Right-Pointing-Pointer The first source of uncertainty is the linear heat rate. Black-Right-Pointing-Pointer The second source of uncertainty is the thermal conductivity of the fuel. - Abstract: In the global framework of nuclear fuel behavior simulation, the response of the models describing the physical phenomena occurring during the irradiation in reactor is mainly conditioned by the confidence in the calculated temperature of the fuel. Amongst all parameters influencing the temperature calculation in our fuel rod simulation code (METEOR V2), several sources of uncertainty have been identified as being the most sensitive: thermal conductivity of UO{sub 2}, radial distribution of power in the fuel pellet, local linear heat rate in the fuel rod, geometry of the pellet and thermal transfer in the gap. Expert judgment and inverse methods have been used to model the uncertainty of these parameters using theoretical distributions and correlation matrices. Propagation of these uncertainties in the METEOR V2 code using the URANIE framework and a Monte-Carlo technique has been performed in different experimental irradiations of UO{sub 2} fuel. At every time step of the simulated experiments, we get a temperature statistical distribution which results from the initial distributions of the uncertain parameters. We then can estimate confidence intervals of the calculated temperature. In order to quantify the sensitivity of the calculated temperature to each of the uncertain input parameters and data, we have also performed a sensitivity analysis using the Sobol' indices at first order.

  4. Advancing the competitive future of nuclear power through partnerships

    International Nuclear Information System (INIS)

    Pryor, Charles W.

    2002-01-01

    Full text: The emergence of competitive markets, coupled with significant environmental benefits, will create a period of tremendous opportunity and enhanced value for commercial nuclear power plants, both existing plants and those on the immediate horizon. The challenge for nuclear plant owners and operators is to reduce costs and improve performance in order to improve the plants net earnings stream and remain competitive with other forms of generation. In his remarks, Dr. Pryor will examine the excellent performance of nuclear plants worldwide and look at the significant value yet to be captured from existing plants. Continued capacity factor improvement driven primarily by shorter outages will help, but new business practices will be needed in the future to achieve competitive production costs of around 1cent/kwh at the bus bar. Dr. Pryor will explore innovative approaches to resource sharing and integration that already are paying dividends worldwide. Beyond this immediate need for greater cooperative efforts is the necessity of longer-range strategies implementing advanced technologies in new plant designs. Although energy demand and environmental benefits are driving renewed interest in nuclear power, the cost challenge for new plants remains significant. Dr. Pryor will discuss the cost challenge, new designs that are available to meet market demand, and the role of utility/supplier partnerships in current and future new plant projects. (author)

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

    International Nuclear Information System (INIS)

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Walter, C.E.

    1999-01-01

    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

  7. Advanced Space Nuclear Reactors from Fiction to Reality

    Science.gov (United States)

    Popa-Simil, L.

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

  8. Issues of high-burnup fuel for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Belac, J.; Milisdoerfer, L.

    2004-12-01

    A brief description is given of nuclear fuels for Generation III+ and IV reactors, and the major steps needed for a successful implementation of new fuels in prospective types of newly designed power reactors are outlined. The following reactor types are discussed: gas cooled fast reactors, heavy metal (lead) cooled fast reactors, molten salt cooled reactors, sodium cooled fast reactors, supercritical water cooled reactors, and very high temperature reactors. The following are regarded as priority areas for future investigations: (i) spent fuel radiotoxicity; (ii) proliferation volatility; (iii) neutron physics characteristics and inherent safety element assessment; technical and economic analysis of the manufacture of advanced fuels; technical and economic analysis of the fuel cycle back end, possibilities of spent nuclear fuel reprocessing, storage and disposal. In parallel, work should be done on the validation and verification of analytical tools using existing and/or newly acquired experimental data. (P.A.)

  9. The 75 years Anniversary of Thermal and Nuclear Energy Department at KTU

    International Nuclear Information System (INIS)

    Gylys, J.

    1997-01-01

    The Thermal and Nuclear Energy Department of Kaunas University of Technology is the only institution educating qualified engineers in thermal and nuclear energy in Lithuania. The first stage of education is a bachelor studies program. The program educates experts for work in thermal and nuclear power plants, steam boiler plants, heat consuming industries, food, chemical, oil processing industries. The bachelors of nuclear engineering are seeking their master degree in the Russian institutes, like Obninsk Institute of Nuclear Power Engineering or in western countries like Sweden and Finland

  10. Nuclear piping criteria for Advanced Light-Water Reactors, Volume 1--Failure mechanisms and corrective actions

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This WRC Bulletin concentrates on the major failure mechanisms observed in nuclear power plant piping during the past three decades and on corrective actions taken to minimize or eliminate such failures. These corrective actions are applicable to both replacement piping and the next generation of light-water reactors. This WRC Bulletin was written with the objective of meeting a need for piping criteria in Advanced Light-Water Reactors, but there is application well beyond the LWR industry. This Volume, in particular, is equally applicable to current nuclear power plants, fossil-fueled power plants, and chemical plants including petrochemical. Implementation of the recommendations for mitigation of specific problems should minimize severe failures or cracking and provide substantial economic benefit. This volume uses a case history approach to high-light various failure mechanisms and the corrective actions used to resolve such failures. Particular attention is given to those mechanisms leading to severe piping failures, where severe denotes complete severance, large ''fishmouth'' failures, or long throughwall cracks releasing a minimum of 50 gpm. The major failure mechanisms causing severe failure are erosion-corrosion and vibrational fatigue. Stress corrosion cracking also has been a common problem in nuclear piping systems. In addition thermal fatigue due to mixing-tee and to thermal stratification also is discussed as is microbiologically-induced corrosion. Finally, water hammer, which represents the ultimate in internally-generated dynamic high-energy loads, is discussed

  11. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2010-01-01

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  12. Advanced passive technology: A global standard for nuclear plant requirements

    International Nuclear Information System (INIS)

    Novak, V.

    1994-01-01

    Since 1984, Westinghouse has been developing AP8OO, a 800 MW, two-loop advanced passive plant, in response to an initiative established by the Electric Power Research Institute (EPRI) and the U.S. Department of Energy' (DOE). The preliminary design was cornpleved in 1989. AP6OO's Standard Safety Analysis and Probabilistic Risk analysis Reports were submitted to the U.S. Nuclear Regulatory Commission for design certification in 1992. Design simplification is the key strategy behind the AP6OO. The basic technical concept Of simplification has resulted in a simplified reactor coolant systems, simplified plant systems, a simplified plant arrangement, reduced number of components, simplified operation and maintenance

  13. Regulatory Risk Management of Advanced Nuclear Power Plants

    International Nuclear Information System (INIS)

    George, Glenn R.

    2002-01-01

    Regulatory risk reflects both the likelihood of adverse outcomes during regulatory interactions and the severity of those outcomes. In the arena of advanced nuclear power plant licensing and construction, such adverse outcomes may include, for example, required design changes and construction delays. These, in turn, could significantly affect the economics of the plant and the generation portfolio in which it will operate. In this paper, the author addresses these issues through the lens of risk management. The paper considers various tools and techniques of regulatory risk management, including design diversity and hedging strategies. The effectiveness of alternate approaches is weighed and recommendations are made in several regulatory contexts. (author)

  14. Nuclear spectroscopy in large shell model spaces: recent advances

    International Nuclear Information System (INIS)

    Kota, V.K.B.

    1995-01-01

    Three different approaches are now available for carrying out nuclear spectroscopy studies in large shell model spaces and they are: (i) the conventional shell model diagonalization approach but taking into account new advances in computer technology; (ii) the recently introduced Monte Carlo method for the shell model; (iii) the spectral averaging theory, based on central limit theorems, in indefinitely large shell model spaces. The various principles, recent applications and possibilities of these three methods are described and the similarity between the Monte Carlo method and the spectral averaging theory is emphasized. (author). 28 refs., 1 fig., 5 tabs

  15. Advanced chemistry management system for nuclear power plants

    International Nuclear Information System (INIS)

    Maeda, Katsuji; Kobayashi, Yasuhiro; Nagasawa, Katsumi

    2000-01-01

    Chemistry control in a boiling water reactor (BWR) plant has a close relationship with radiation field buildup, fuel reliability, integrity of plant components and materials, performance of the water treatment systems and radioactive waste generation. Chemistry management in BWR plants has become more important in order to maintain and enhance plant reliability. Adequate chemistry control and management are also essential to establish, maintain, and enhance plant availability. For these reasons, we have developed the advanced chemistry management system for nuclear power plants in order to effectively collect and evaluate a large number of plant operating and chemistry data. (author)

  16. Testing Systems and Results for Advanced Nuclear Fuel Materials

    International Nuclear Information System (INIS)

    Rooyen, I.J. van; Griffith, G.W.; Garnier, J.E.

    2012-01-01

    Light Water Reactor Sustainability (LWRS) Program Advanced LWR Nuclear Fuel Development (ALFD) Pathway. Development and testing of high performance fuel cladding identified as high priority to support: enhancement of fuel performance, reliability, and reactor safety. One of the technologies being examined is an advanced fuel cladding made from ceramic matrix composites (CMC) utilizing silicon carbide (SiC) as a structural material supplementing a commercial Zircaloy-4 (Zr-4) tube. A series of out-of-pile tests to fully characterize the SiC CMC hybrid design to produce baseline data. The planned tests are intended to either produce quantitative data or to demonstrate the properties required to achieve two initial performance conditions relative to standard zircaloybased cladding: decreased hydrogen uptake (corrosion) and decreased fretting of the cladding tube under normal operating and postulated accident conditions. These two failure mechanisms account for approximately 70% of all in-pile failures of LWR commercial fuel assemblies

  17. Advanced man-machine system for nuclear power plants

    International Nuclear Information System (INIS)

    Masui, Takao; Naito, Norio; Kato, Kanji.

    1990-01-01

    Recent development of artificial intelligence(AI) seems to offer new possibility to strengthen the performance of the operator support system. From this point of view, a national project of Advanced Man-Machine System Development for Nuclear Power Plant (MMS-NPP) has been carried out since 1984 as 8-year project. This project aims at establishing advanced operator support functions which support operators in their knowledge-based behaviors and smoother interface with the system. This paper describes the role of MMS-NPP, the support functions and the main feature of the MMS-NPP detailed design with its focus placed on the realization methods using AI technology of the support functions for BWR and PWR plants. (author)

  18. Status Report on Structural Materials for Advanced Nuclear Systems

    International Nuclear Information System (INIS)

    Allen, T.R.; Balbaud-Celerier, F.; Asayama, T.; Pouchon, M.; Busby, J.T.; Maloy, S.; Park, J.Y.; Fazio, C.; Dai, Y.; Agostini, P.; Chevalier, J.P.; Marrow, J.

    2013-01-01

    Materials performance is critical to the safe and economic operation of any nuclear system. As the international community pursues the development of Generation IV reactor concepts and accelerator-driven transmutation systems, it will be increasingly necessary to develop advanced materials capable of tolerating the more challenging environments of these new systems. The international community supports numerous materials research programmes, with each country determining its individual focus on a case-by-case basis. In many instances, similar alloys of materials systems are being studied in several countries, providing the opportunity for collaborative and cross-cutting research that benefits different systems. This report is a snapshot of the current materials programmes supporting the development of advanced concepts. The descriptions of the research are grouped by concept, and national programmes are described within each concept. The report provides an overall sense of the importance of materials research worldwide and the opportunities for synergy among the countries represented in this overview. (authors)

  19. National project : advanced robot for nuclear power plant

    International Nuclear Information System (INIS)

    Tsunemi, T.; Takehara, K.; Hayashi, T.; Okano, H.; Sugiyama, S.

    1993-01-01

    The national project 'Advanced Robot' has been promoted by the Agency of Industrial science and Technology, MITI for eight years since 1983. The robot for a nuclear plant is one of the projects, and is a prototype intelligent one that also has a three dimensional vision system to generate an environmental model, a quadrupedal walking mechanism to work on stairs and four fingered manipulators to disassemble a valve with a hand tool. Many basic technologies such as an actuator, a tactile sensor, autonomous control and so on progress to high level. The prototype robot succeeded functionally in official demonstration in 1990. More refining such as downsizing and higher intelligence is necessary to realize a commercial robot, while basic technologies are useful to improve conventional robots and systems. This paper presents application studies on the advanced robot technologies. (author)

  20. Advanced Instrumentation and control techniques for nuclear power plants

    International Nuclear Information System (INIS)

    Mori, Nobuyuki; Makino, Maomi; Naito, Norio

    1992-01-01

    Toshiba has been promoting the development of an advanced instrumentation and control system for nuclear power plants to fulfill the requirements for increased reliability, improved functionality and maintainability, and more competitive economic performance. This system integrates state-of-the-art technologies such as those for the latest man-machine interface, digital processing, optical multiplexing signal transmission, human engineering, and artificial intelligence. Such development has been systematically accomplished based on a schematic view of integrated digital control and instrumentation systems, and the development of whole systems has now been completed. This paper describes the purpose, design philosophy, and contents of newly developed systems, then considers the future trends of advanced man-machine systems. (author)

  1. Method for limiting movement of a thermal shield for a nuclear reactor, and thermal shield displacement limiter therefor

    International Nuclear Information System (INIS)

    Meuschke, R.E.; Boyd, C.H.

    1989-01-01

    This patent describes a method of limiting the movement of a thermal shield of a nuclear reactor. It comprises: machining at least four (4) pockets in upper portions of a thermal shield circumferentially about a core barrel of a nuclear reactor to receive key-wave inserts; tapping bolt holes in the pockets of the thermal shield to receive bolts; positioning key-wave inserts into the pockets of the thermal shield to be bolted in place with the bolt holes; machining dowel holes at least partially through the positioned key-way inserts and the thermal shield to receive dowel pins; positioning dowel pins in the dowel holes in the key-way insert and thermal shield to tangentially restrain movement of the thermal shield relative to the core barrel; sliding limiter keys into the key-way inserts and bolting the limiter keys to the core barrel to tangentially restrain movement of the thermal shield relative and the core barrel while allowing radial and axial movement of the thermal shield relative to the core barrel; machining dowel holes through the limiter key and at least partially through the core barrel to receive dowel pins; positioning dowel pins in the dowel holes in the limiter key and core barrel to restrain tangential movement of the thermal shield relative to the core barrel of the nuclear reactor

  2. Review of Nuclear Thermal Propulsion Ground Test Options

    Science.gov (United States)

    Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

    2015-01-01

    High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

  3. Potential impact of thermal effluents from Chongqing Fuling nuclear power plant to the Three Gorges Reservoir

    International Nuclear Information System (INIS)

    Han Baohua; Li Jianguo; Ma Binghui; Zhang Yue; Sun Qunli; Hu Yuping

    2012-01-01

    This study is based on the hydrological data near Chongqing Fuling Nuclear Power Plant along the Yangtze River, the present situation of the ecological environment of the Three Gorges Reservoir and the predicted results of thermal effluents from Chongqing Fuling Nuclear Power Plant. The standards of cooling water and the thermal tolerances indexes of aquatic organisms were investigated. The effects of thermal effluents on aquatic organisms were analyzed. The potential impact of Chongqing Fuling nuclear power plant to the Three Gorges Reservoir was explained. The results show that in the most adverse working conditions, the surface temperature near the outfall area is not more than 1℃, the temperature of thermal effluents do not exceed the suitable thermal range of fish breeding, growth and other thermal tolerances indexes. Thermal effluents from nuclear power plant have no influence about fish, plankton and benthic organisms in the Three Gorges Reservoir. (authors)

  4. Advanced digital technology - improving nuclear power plant performance through maintainability

    International Nuclear Information System (INIS)

    Ford, J.L.; Senechal, R.R.; Altenhein, G.D.; Harvey, R.P.

    1998-01-01

    In today's energy sector there is ever increasing pressure on utilities to operate power plants at high capacity factors. To ensure nuclear power is competitive into the next century, it is imperative that strategic design improvements be made to enhance the performance of nuclear power plants. There are a number of factors that affect a nuclear power plant's performance; lifetime maintenance is one of the major contributors. The maturing of digital technology has afforded ABB the opportunity to make significant design improvements in the area of maintainability. In keeping with ABB's evolutionary advanced nuclear plant design approach, digital technology has systematically been incorporated into the control and protection systems of the most recent Korean nuclear units in operation and under construction. One example of this was the multi-functional design team approach that was utilized for the development of ABB's Digital Plant Protection System (DPPS). The design team consisted of engineers, maintenance technicians, procurement specialists and manufacturing personnel in order to provide a complete perspective on all facets of the design. The governing design goals of increased reliability and safety, simplicity of design, use of off-the-shelf products and reduced need for periodic surveillance testing were met with the selection of proven ABB-Advant Programmable Logic Controllers (PLCs) as the heart of the DPPS. The application of digital PLC technology allows operation for extended periods without requiring routine maintenance or re-calibration. A well documented commercial dedication program approved by the United States Nuclear Regulatory Commission (US NRC) as part of the System 80+ TM Advanced Light Water Reactor Design Certification Program, allowed the use of off-the shelf products in the design of the safety protection system. In addition, a number of mechanical and electrical improvements were made which support maintainability. The result is a DPPS

  5. Applicability of advanced automotive heat engines to solar thermal power

    Science.gov (United States)

    Beremand, D. G.; Evans, D. G.; Alger, D. L.

    The requirements of a solar thermal power system are reviewed and compared with the predicted characteristics of automobile engines under development. A good match is found in terms of power level and efficiency when the automobile engines, designed for maximum powers of 65-100 kW (87 to 133 hp) are operated to the nominal 20-40 kW electric output requirement of the solar thermal application. At these reduced power levels it appears that the automotive gas turbine and Stirling engines have the potential to deliver the 40+ percent efficiency goal of the solar thermal program.

  6. Thermal Evaluation of Storage Rack with an Advanced Neutron Absorber during Normal Operation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hee-Jae; Kim, Mi-Jin; Sohn, Dong-Seong [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    The storage capacity of the domestic wet storage site is expected to reach saturation from Hanbit in 2024 to Sin-wolseong in 2038 and accordingly management alternatives are urgently taken. Since installation of the dense rack is considered in the short term, it is necessary to urgently develop an advanced neutron absorber which can be applied to a spent nuclear fuel storage facility. Neutron absorber is the material for controlling the reactivity. A material which has excellent thermal neutron absorption ability, high strength and corrosion resistance must be selected as the neutron absorber. Existing neutron absorbers are made of boron which has a good thermal absorption ability such as BORAL and METAMIC. However, possible problems have been reported in using the boron-based neutron absorber for wet storage facility. Gadolinium is known to have higher neutron absorption cross-section than that of boron. And the strength of duplex stainless steel is about 1.5 times higher than stainless steel 304 which has been frequently used as a structural material. Therefore, duplex stainless steel which contains gadolinium is in consideration as an advanced neutron absorber. Temperature distribution is shown in figure 4. In pool bottom region near the inlet shows a relatively low tendency and heat generated from the fuel assemblies is transmitted to the pool upper region by the vertical flow. Also, temperature gradient appear in rack structures for the axial direction and temperature is uniformly distributed in the pool upper region. Table 1 presents the calculated results. The maximum temperature is 306.63K and does not exceed the 333.15K (60℃). The maximum temperature of the neutron absorber is 306.48K.

  7. Thermal pollution of rivers and reservoirs by discharges of heated water from thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Makarov, I.

    1974-12-01

    The problems are discussed of the thermal pollution of rivers and water reservoirs by discharges of heated water from thermal and nuclear power plants. The problems concerned are quantitative and qualitative changes in biocenoses, the disturbance or extinction of flora and fauna, physiological changes in organisms and changes in the hydrochemical regime. (Z.M.)

  8. Diagnosis of Thermal Efficiency of Nuclear Power Plants Using Optical Torque Sensors

    International Nuclear Information System (INIS)

    Shuichi Umezawa; Jun Adachi

    2006-01-01

    A new optical torque measuring method was applied to diagnosis of thermal efficiency of nuclear power plants. The sensor allows torque deformation of the rotor caused by power transmission to be measured without contact. Semiconductor laser beams and small pieces of stainless reflector that have bar-code patterns are employed. The intensity of the reflected laser beam is measured and then input into a computer through an APD and an A/D converter having high frequency sampling rates. The correlation analysis technique can translate these data into the torque deformation angle. This angle allows us to obtain the turbine output along with the torsional rigidity and the rotating speed of the rotor. The sensor was applied to a nuclear plant of Tokyo Electric Power Company, TEPCO, following its application success to the early combined cycle plants and the advanced combined cycle plants of TEPCO. As the turbine rotor of the nuclear power plant is less exposed than that of the combined cycle plants, the measurement position is confined to a narrow gap. In order to overcome the difficulty in installation, the shape of the sensor is modified to be long and thin. Sensor performance of the nuclear power plant was inspected over a year. The value of the torsional rigidity was analyzed by the finite element method at first. Accuracy was improved by correcting the torsional rigidity so that the value was consistent with the generator output. As a result, it is considered that the sensor performance has reached a practical use level. (authors)

  9. Research on Elemental Technology of Advanced Nuclear Fuel Performance Verification

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Dong Uk; Jean, Sang Hwan; Koo, Min

    2003-04-01

    Most of current properties models and fuel performance models used in the performance evaluation codes are based on the in-pile data up to 33,000 MWd/MtU. Therefore, international experts are investigating the properties changes and developing advanced prediction models for high burn-up application. Current research is to develop high burn-up fission gas release model for the code and to support the code development activities by collecting data and models, reviewing/assessing the data and models together, and benchmarking the selected models against the appropriate in-pile data. For high burn-up applications, two stage two step fission gas release model is developed based on the real two diffusion process in the grain lattice and grain boundaries of the fission gases and the observation of accelerated release rate in the high burn-up. It is found that the prediction of this model is in excellent agreement with the in-pile measurement results, not only in the low burn-up but also in the high burn-up. This research is found that the importance of thermal conductivity of oxide fuel, especially in the high burn-up, is focused again. It is found that even the temperature dependent models differ from one to another and most of them overestimate the conductivity in the high burn-up. An in-pile data benchmarking of high LHGR fuel rod shows that the difference can reach 30%∼40%, which predicts 400 .deg. C lower than the real fuel centerline temperature. Recent models on the thermal expansion and heat capacity of oxide fuel are found to be well-defined. Irradiation swelling of the oxide fuel are now well-understood that in most cases in LWRs solid fission product swelling is dominant. Thus, the accumulation of in-pile data can enhance the accuracy of the model prediction, rather than theoretical modeling works. Thermo-physical properties of Zircaloy cladding are also well-defined and well-understood except the thermal expansion. However, it turns out that even the

  10. Contributions to thermal and fluid dynamic problems in nuclear technology

    International Nuclear Information System (INIS)

    Mueller, U.; Krebs, L.; Rust, K.

    1984-02-01

    The majority of contributions compiled in this report deals with thermal and fluid dynamic problems in nuclear engineering. Especially problems of heat transfer and cooling are represented which may arise during and afer a loss-of-coolant accident both in light water reactors and in liquid metal cooled fast breeder reactors. Papers on the mass transfer in pressurized water, tribological problems in sodium cooled reactors, the fluid dynamics of pulsed column, and fundamental investigations of convective flows supplement these contributions on problems connected with accidents. Furthermore, a keynote paper presents the individual activities relating to the reliability of reactor components, a field recently included in our research program. Technical solutions to special problems are closely connected to the investigations based on experiments. Therefore, several contributions deal with new developments in technology and measuring techniques. (orig.) [de

  11. Thermal hydraulic analysis of the encapsulated nuclear heat source

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Wade, D.C. [Argonne National Lab., IL (United States)

    2001-07-01

    An analysis has been carried out of the steady state thermal hydraulic performance of the Encapsulated Nuclear Heat Source (ENHS) 125 MWt, heavy liquid metal coolant (HLMC) reactor concept at nominal operating power and shutdown decay heat levels. The analysis includes the development and application of correlation-type analytical solutions based upon first principles modeling of the ENHS concept that encompass both pure as well as gas injection augmented natural circulation conditions, and primary-to-intermediate coolant heat transfer. The results indicate that natural circulation of the primary coolant is effective in removing heat from the core and transferring it to the intermediate coolant without the attainment of excessive coolant temperatures. (authors)

  12. Tools for advanced simulations to nuclear propulsion systems in rockets

    International Nuclear Information System (INIS)

    Torres Sepulveda, A.; Perez Vara, R.

    2004-01-01

    While chemical propulsion rockets have dominated space exploration, other forms of rocket propulsion based on nuclear power, electrostatic and magnetic drive, and other principles besides chemical reactions, have been considered from the earliest days of the field. The goal of most of these advanced rocket propulsion schemes is improved efficiency through higher exhaust velocities, in order to reduce the amount of fuel the rocket vehicle needs to carry, though generally at the expense of high thrust. Nuclear propulsion seems to be the most promising short term technology to plan realistic interplanetary missions. The development of a nuclear electric propulsion spacecraft shall require the development of models to analyse the mission and to understand the interaction between the related subsystems (nuclear reactor, electrical converter, power management and distribution, and electric propulsion) during the different phases of the mission. This paper explores the modelling of a nuclear electric propulsion (NEP) spacecraft type using EcosimPro simulation software. This software is a multi-disciplinary simulation tool with a powerful object-oriented simulation language and state-of-the-art solvers. EcosimPro is the recommended ESA simulation tool for environmental Control and Life Support Systems (ECLSS) and has been used successfully within the framework of the European activities of the International Space Station programme. Furthermore, propulsion libraries for chemical and electrical propulsion are currently being developed under ESA contracts to set this tool as standard usage in the propulsion community. At present, there is not any workable NEP spacecraft, but a standardized-modular, multi-purpose interplanetary spacecraft for post-2000 missions, called ISC-2000, has been proposed in reference. The simulation model presented on this paper is based on the preliminary designs for this spacecraft. (Author)

  13. Advanced Thermal Interface Material Systems for Space Applications, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The ultimate aim of proposed efforts are to develop innovative material and process (M increase thermal cycles before degradation and efforts to ensure ease of...

  14. Integrated System Modeling for Nuclear Thermal Propulsion (NTP)

    Science.gov (United States)

    Ryan, Stephen W.; Borowski, Stanley K.

    2014-01-01

    Nuclear thermal propulsion (NTP) has long been identified as a key enabling technology for space exploration beyond LEO. From Wernher Von Braun's early concepts for crewed missions to the Moon and Mars to the current Mars Design Reference Architecture (DRA) 5.0 and recent lunar and asteroid mission studies, the high thrust and specific impulse of NTP opens up possibilities such as reusability that are just not feasible with competing approaches. Although NTP technology was proven in the Rover / NERVA projects in the early days of the space program, an integrated spacecraft using NTP has never been developed. Such a spacecraft presents a challenging multidisciplinary systems integration problem. The disciplines that must come together include not only nuclear propulsion and power, but also thermal management, power, structures, orbital dynamics, etc. Some of this integration logic was incorporated into a vehicle sizing code developed at NASA's Glenn Research Center (GRC) in the early 1990s called MOMMA, and later into an Excel-based tool called SIZER. Recently, a team at GRC has developed an open source framework for solving Multidisciplinary Design, Analysis and Optimization (MDAO) problems called OpenMDAO. A modeling approach is presented that builds on previous work in NTP vehicle sizing and mission analysis by making use of the OpenMDAO framework to enable modular and reconfigurable representations of various NTP vehicle configurations and mission scenarios. This approach is currently applied to vehicle sizing, but is extensible to optimization of vehicle and mission designs. The key features of the code will be discussed and examples of NTP transfer vehicles and candidate missions will be presented.

  15. Center of thermal-physical data for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bobkov, V P; Blokhin, A I; Ivashkevich, A A; Katan, I B; Peskov, O L; Pan' kov, V M; Savanin, N K; Sal' nikova, O V; Khrushcheva, E N; Kirova, T S

    1982-09-01

    The specific features of a specialized Center of thermal-physical data (CTD) are considered. The center has been created for data acquisition, storage and analysis and working out recommendations on the following NPP thermal physics sections: hydrodynamics of channel flows (monophase laminar and turbulent, and two-phase flows, hydrodynamic vibrations) heat exchange in NPP elements, thermohydraulic calculations of nuclear reactor cores, heat exchangers, steam generators and NPP cooling system elements, coolant properties (water and steam, liquid metals and gases). On the CTD data base an automated system ASKhOD, oriented to EC computer, is created. The ASKhoD software ensures data allocation on magnetic tapes or other carriers, automated renewal and data relocation, data search in compliance with a specified set of signs, data processing for the purpose of their estimation or obtaining optimized model constants. Different publications in home and foreign magazines, conference, seminar materials, organization preprints serve as the data sources used for the formation of the ASKhOD data base.

  16. Center of thermal-physical data for nuclear power plants

    International Nuclear Information System (INIS)

    Bobkov, V.P.; Blokhin, A.I.; Ivashkevich, A.A.; Katan, I.B.; Peskov, O.L.; Pan'kov, V.M.; Savanin, N.K.; Sal'nikova, O.V.; Khrushcheva, E.N.; Kirova, T.S.

    1982-01-01

    The specific features of a specialized Center of thermal-physical data (CTD) are considered. The center has been created for data acquisition, storage and analysis and working out recommendations on the following NPP thermal physics sections: hydrodynamics of channel flows (monophase laminar and turbulent, and two-phase flows, hydrodynamic vibrations) heat exchange in NPP elements, thermohydraulic calculations of nuclear reactor cores, heat exchangers, steam generators and NPP cooling system elements, coolant properties (water and steam, liquid metals and gases). On the CTD data base an automated system ASKhOD, oriented to EC computer, is created. The ASKhoD software ensures data allocation on magnetic tapes or other carriers, automated renewal and data relocation, data search in compliance with a specified set of signs, data processing for the purpose of their estimation or obtaining optimized model constants. Different publications in home and foreign magazines, conference, seminar materials, organization preprints serve as the data sources used for the formation of the ASKhOD data base

  17. Nuclear thermal rocket propulsion application to Mars missions

    International Nuclear Information System (INIS)

    Emrich, W.J. Jr.; Young, A.C.; Mulqueen, J.A.

    1991-01-01

    Options for vehicle configurations are reviewed in which nuclear thermal rocket (NTR) propulsion is used for a reference mission to Mars. The scenario assumes an opposition-class Mars transfer trajectory, a 435-day mission, and the use of a single nuclear engine with 75,000 lbs of thrust. Engine parameters are examined by calculating mission variables for a range of specific impulses and thrust/weight ratios. The reference mission is found to have optimal values of 925 s for the specific impulse and thrust/weight ratios of 4.0 and 0.06 for the engine and total stage ratios respectively. When the engine thrust/weight ratio is at least 4/1 the most critical engine parameter is engine specific impulse for reducing overall stage weight. In the context of this trans-Mars three-burn maneuver the NTR engine with an expander engine cycle is considered a more effective alternative than chemical/aerobrake and other propulsion options

  18. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  19. Comparative analysis of thermal behavior in hollow nuclear fuel pellets

    International Nuclear Information System (INIS)

    Santos, Beatriz M. dos; Alvim, Antonio C.M.

    2017-01-01

    The increase in energy demand in Brazil and in the world is a real problem and several solutions are being considered to mitigate it. Maximization of energy generation, within the safety standards of fuel resources already known, is one of them. In this respect, nuclear energy is a crucial technology to sustain energy demand on several countries. Performances of a solid cylindrical and an annular rod have been verified and compared; where it has been proven that the annular rod can reach a higher nominal power in relation to the solid one. In this paper, the temperature profiles of two distinct nuclear fuel pellets, one of them annular and the other in the shape of a hollow biconcave disc (like the cross section of a red blood cell), were compared to analyze the efficiency and safety of both. The finite differences method allowed the evaluation of the thermal behavior of these pellets, where one specific physical condition was analyzed, regarding convection and conduction at the lateral edges. The results show that the temperature profile of the hollow biconcave disc pellet is lower, about 70 deg C below, when compared to the temperature profile of the annular pellet, considering the same simulation parameters for both pellets. (author)

  20. Comparative analysis of thermal behavior in hollow nuclear fuel pellets

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Beatriz M. dos; Alvim, Antonio C.M., E-mail: bmachado@nuclear.ufrj.br, E-mail: aalvim@gmail.com [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-11-01

    The increase in energy demand in Brazil and in the world is a real problem and several solutions are being considered to mitigate it. Maximization of energy generation, within the safety standards of fuel resources already known, is one of them. In this respect, nuclear energy is a crucial technology to sustain energy demand on several countries. Performances of a solid cylindrical and an annular rod have been verified and compared; where it has been proven that the annular rod can reach a higher nominal power in relation to the solid one. In this paper, the temperature profiles of two distinct nuclear fuel pellets, one of them annular and the other in the shape of a hollow biconcave disc (like the cross section of a red blood cell), were compared to analyze the efficiency and safety of both. The finite differences method allowed the evaluation of the thermal behavior of these pellets, where one specific physical condition was analyzed, regarding convection and conduction at the lateral edges. The results show that the temperature profile of the hollow biconcave disc pellet is lower, about 70 deg C below, when compared to the temperature profile of the annular pellet, considering the same simulation parameters for both pellets. (author)

  1. High-temperature turbopump assembly for space nuclear thermal propulsion

    Science.gov (United States)

    Overholt, David M.

    1993-01-01

    The development of a practical, high-performance nuclear rocket by the U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) program places high priority on maximizing specific impulse (ISP) and thrust-to-weight ratio. The operating parameters arising from these goals drive the propellant-pump design. The liquid hydrogen propellant is pressurized and pumped to the reactor inlet by the turbopump assembly (TPA). Rocket propulsion is effected by rapid heating of the propellant from 100 K to thousands of degrees in the particle-bed reactor (PBR). The exhausted propellant is then expanded through a high-temperature nozzle. One approach to achieve high performance is to use an uncooled carbon-carbon nozzle and duct turbine inlet. The high-temperature capability is obtained by using carbon-carbon throughout the TPA hot section. Carbon-carbon components in development include structural parts, turbine nozzles/stators, and turbine rotors. The technology spinoff is applicable to conventional liquid propulsion engines plus a wide variety of other turbomachinery applications.

  2. Carbon-carbon turbopump concept for Space Nuclear Thermal Propulsion

    Science.gov (United States)

    Overholt, David M.

    1993-06-01

    The U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) program is placing high priority on maximizing specific impulse (ISP) and thrust-to-weight ratio in the development of a practical high-performance nuclear rocket. The turbopump design is driven by these goals. The liquid hydrogen propellant is pressurized and pumped to the reactor inlet by the turbopump assembly (TPA). Rocket propulsion is from rapid heating of the propellant from 180 R to thousands of degrees in the particle bed reactor (PBR). The exhausted propellant is then expanded through a high-temperature nozzle. A high-performance approach is to use an uncooled carbon-carbon nozzle and duct turbine inlet. Carbon-carbon components are used throughout the TPA hot section to obtain the high-temperature capability. Several carbon-carbon components are in development including structural parts, turbine nozzles/stators, and turbine rotors. The technology spinoff is applicable to conventional liquid propulsion engines and many other turbomachinery applications.

  3. High-temperature turbopump assembly for space nuclear thermal propulsion

    International Nuclear Information System (INIS)

    Overholt, D.M.

    1993-01-01

    The development of a practical, high-performance nuclear rocket by the U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) program places high priority on maximizing specific impulse (ISP) and thrust-to-weight ratio. The operating parameters arising from these goals drive the propellant-pump design. The liquid hydrogen propellant is pressurized and pumped to the reactor inlet by the turbopump assembly (TPA). Rocket propulsion is effected by rapid heating of the propellant from 100 K to thousands of degrees in the particle-bed reactor (PBR). The exhausted propellant is then expanded through a high-temperature nozzle. One approach to achieve high performance is to use an uncooled carbon-carbon nozzle and duct turbine inlet. The high-temperature capability is obtained by using carbon-carbon throughout the TPA hot section. Carbon-carbon components in development include structural parts, turbine nozzles/stators, and turbine rotors. The technology spinoff is applicable to conventional liquid propulsion engines plus a wide variety of other turbomachinery applications

  4. Carbon-carbon turbopump concept for Space Nuclear Thermal Propulsion

    International Nuclear Information System (INIS)

    Overholt, D.M.

    1993-06-01

    The U.S. Air Force Space Nuclear Thermal Propulsion (SNTP) program is placing high priority on maximizing specific impulse (ISP) and thrust-to-weight ratio in the development of a practical high-performance nuclear rocket. The turbopump design is driven by these goals. The liquid hydrogen propellant is pressurized and pumped to the reactor inlet by the turbopump assembly (TPA). Rocket propulsion is from rapid heating of the propellant from 180 R to thousands of degrees in the particle bed reactor (PBR). The exhausted propellant is then expanded through a high-temperature nozzle. A high-performance approach is to use an uncooled carbon-carbon nozzle and duct turbine inlet. Carbon-carbon components are used throughout the TPA hot section to obtain the high-temperature capability. Several carbon-carbon components are in development including structural parts, turbine nozzles/stators, and turbine rotors. The technology spinoff is applicable to conventional liquid propulsion engines and many other turbomachinery applications. 3 refs

  5. 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.

  6. 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.

  7. Pressurized thermal shock analysis in German nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Fricke, Stefan; Braun, Michael [TUEV NORD Nuclear, Hannover (Germany)

    2015-03-15

    For more than 30 years TUeV NORD is a competent consultant in nuclear safety is-sues giving expert third party opinion to our clients. According to the German regulations the safety against brittle fracture has to be proved for the reactor pressure vessel (RPV) and with a new level of knowledge the proof has to be continuously updated with the development in international codes and standards like ASME, BS and RCC-M. The load of the RPV is a very complex transient pressure and temperature situation. Today these loading conditions can be modeled by thermal hydraulic calculations and new experimental results much more detailed than in the construction phase of German Nuclear Power Plants in the 1980s. Therefore, the proof against brittle fracture from the construction phase had to be updated for all German Nuclear Power Plants with the new findings of the loading conditions especially for a postulated small leakage in the main coolant line. The RPV consists of ferritic base material (about 250 mm) and austenitic cladding (about 6 mm) at the inner side. The base material and the cladding have different physical properties which have to be considered temperature dependently in the cal-culations. Radiation-embrittlement effects on the material are to be respected in the fracture mechanics assessment. The regions of the RPV of special interest are the core weld, the inlet and outlet nozzle region and the flange connecting weld zone. The fracture mechanics assessment is performed for normal and abnormal operating conditions and for accidents like LOCA (Loss of Coolant Accident). In this paper the German approach to fracture mechanics assessment to brittle fracture will be discussed from the point of view of a third party organization.

  8. Application of monitoring, diagnosis, and prognosis in thermal performance analysis for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Min; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2014-12-15

    As condition-based maintenance (CBM) has risen as a new trend, there has been an active movement to apply information technology for effective implementation of CBM in power plants. This motivation is widespread in operations and maintenance, including monitoring, diagnosis, prognosis, and decision-making on asset management. Thermal efficiency analysis in nuclear power plants (NPPs) is a longstanding concern being updated with new methodologies in an advanced IT environment. It is also a prominent way to differentiate competitiveness in terms of operations and maintenance costs. Although thermal performance tests implemented using industrial codes and standards can provide officially trustworthy results, they are essentially resource-consuming and maybe even a hind-sighted technique rather than a foresighted one, considering their periodicity. Therefore, if more accurate performance monitoring can be achieved using advanced data analysis techniques, we can expect more optimized operations and maintenance. This paper proposes a framework and describes associated methodologies for in-situ thermal performance analysis, which differs from conventional performance monitoring. The methodologies are effective for monitoring, diagnosis, and prognosis in pursuit of CBM. Our enabling techniques cover the intelligent removal of random and systematic errors, deviation detection between a best condition and a currently measured condition, degradation diagnosis using a structured knowledge base, and prognosis for decision-making about maintenance tasks. We also discuss how our new methods can be incorporated with existing performance tests. We provide guidance and directions for developers and end-users interested in in-situ thermal performance management, particularly in NPPs with large steam turbines.

  9. Application of monitoring, diagnosis, and prognosis in thermal performance analysis for nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Hyeong Min; Heo, Gyun Young; Na, Man Gyun

    2014-01-01

    As condition-based maintenance (CBM) has risen as a new trend, there has been an active movement to apply information technology for effective implementation of CBM in power plants. This motivation is widespread in operations and maintenance, including monitoring, diagnosis, prognosis, and decision-making on asset management. Thermal efficiency analysis in nuclear power plants (NPPs) is a longstanding concern being updated with new methodologies in an advanced IT environment. It is also a prominent way to differentiate competitiveness in terms of operations and maintenance costs. Although thermal performance tests implemented using industrial codes and standards can provide officially trustworthy results, they are essentially resource-consuming and maybe even a hind-sighted technique rather than a foresighted one, considering their periodicity. Therefore, if more accurate performance monitoring can be achieved using advanced data analysis techniques, we can expect more optimized operations and maintenance. This paper proposes a framework and describes associated methodologies for in-situ thermal performance analysis, which differs from conventional performance monitoring. The methodologies are effective for monitoring, diagnosis, and prognosis in pursuit of CBM. Our enabling techniques cover the intelligent removal of random and systematic errors, deviation detection between a best condition and a currently measured condition, degradation diagnosis using a structured knowledge base, and prognosis for decision-making about maintenance tasks. We also discuss how our new methods can be incorporated with existing performance tests. We provide guidance and directions for developers and end-users interested in in-situ thermal performance management, particularly in NPPs with large steam turbines.

  10. Relevant thermal hydraulic aspects of advanced reactors design: status report

    International Nuclear Information System (INIS)

    1996-11-01

    This status report provides an overview on the relevant thermalhydraulic aspects of advanced reactor designs (e.g. ABWR, AP600, SBWR, EPR, ABB 80+, PIUS, etc.). Since all of the advanced reactor concepts are at the design stage, the information and data available in the open literature are still very limited. Some characteristics of advanced reactor designs are provided together with selected phenomena identification and ranking tables. Specific needs for thermalhydraulic codes together with the list of relevant and important thermalhydraulic phenomena for advanced reactor designs are summarized with the purpose of providing some guidance in development of research plans for considering further code development and assessment needs and for the planning of experimental programs

  11. Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options

    International Nuclear Information System (INIS)

    Brent W. Dixon; Steven J. Piet

    2004-01-01

    tripling market share by 2100 from the current 8.4% to 25%, equivalent to continuing the average market growth of last 50 years for an additional 100 years. Five primary spent fuel management strategies are assessed against each of the energy futures to determine the number of geological repositories needed and how the first repository would be used. The geological repository site at Yucca Mountain, Nevada, has the physical potential to accommodate all the spent fuel that will be generated by the current fleet of domestic commercial nuclear reactors, even with license extensions. If new nuclear plants are built in the future as replacements or additions, the United States will need to adopt spent fuel treatment to extend the life of the repository. Should a significant number of new nuclear plants be built, advanced fuel recycling will be needed to fully manage the spent fuel within a single repository. The analysis also considers the timeframe for most efficient implementation of new spent fuel management strategies. The mix of unprocessed spent fuel and processed high level waste in Yucca Mountain varies with each future and strategy. Either recycling must start before there is too much unprocessed waste emplaced or unprocessed waste will have to be retrieved later with corresponding costs. For each case, the latest date to implement reprocessing without subsequent retrieval is determined

  12. A brief review of advances in complex networks of nuclear science and technology field

    International Nuclear Information System (INIS)

    Fang Jinqing

    2010-01-01

    A brief review of advances in complex networks of nuclear science and technology field at home and is given and summarized. These complex networks include: nuclear energy weapon network, network centric warfare, beam transport networks, continuum percolation evolving network associated with nuclear reactions, global nuclear power station network, (nuclear) chemistry reaction networks, radiological monitoring and anti-nuclear terror networks, and so on. Some challenge issues and development prospects of network science are pointed out finally. (authors)

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

    International Nuclear Information System (INIS)

    Golay, M.W.

    1993-01-01

    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. Efficiency improvement of nuclear power plant operation: the significant role of advanced nuclear fuel technologies

    International Nuclear Information System (INIS)

    Velde Van de, A.; Burtak, F.

    2001-01-01

    Due to the increased liberalisation of the power markets, nuclear power generation is being exposed to high cost reduction pressure. In this paper we highlight the role of advanced nuclear fuel technologies to reduce the fuel cycle costs and therefore increase the efficiency of nuclear power plant operation. The key factor is a more efficient utilisation of the fuel and present developments at Siemens are consequently directed at (i) further increase of batch average burnup, (ii) improvement of fuel reliability, (iii) enlargement of fuel operation margins and (iv) improvement of methods for fuel design and core analysis. As a result, the nuclear fuel cycle costs for a typical LWR have been reduced during the past decades by about US$ 35 million per year. The estimated impact of further burnup increases on the fuel cycle costs is expected to be an additional saving of US$10 - 15 million per year. Due to the fact that the fuel will operate closer to design limits, a careful approach is required when introducing advanced fuel features in reload quantities. Trust and co-operation between the fuel vendors and the utilities is a prerequisite for the common success. (authors)

  15. Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

    Science.gov (United States)

    Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

    2004-02-01

    A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

  16. 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

  17. Advanced Signal Processing for Thermal Flaw Detection; TOPICAL

    International Nuclear Information System (INIS)

    VALLEY, MICHAEL T.; HANSCHE, BRUCE D.; PAEZ, THOMAS L.; URBINA, ANGEL; ASHBAUGH, DENNIS M.

    2001-01-01

    Dynamic thermography is a promising technology for inspecting metallic and composite structures used in high-consequence industries. However, the reliability and inspection sensitivity of this technology has historically been limited by the need for extensive operator experience and the use of human judgment and visual acuity to detect flaws in the large volume of infrared image data collected. To overcome these limitations new automated data analysis algorithms and software is needed. The primary objectives of this research effort were to develop a data processing methodology that is tied to the underlying physics, which reduces or removes the data interpretation requirements, and which eliminates the need to look at significant numbers of data frames to determine if a flaw is present. Considering the strengths and weakness of previous research efforts, this research elected to couple both the temporal and spatial attributes of the surface temperature. Of the possible algorithms investigated, the best performing was a radiance weighted root mean square Laplacian metric that included a multiplicative surface effect correction factor and a novel spatio-temporal parametric model for data smoothing. This metric demonstrated the potential for detecting flaws smaller than 0.075 inch in inspection areas on the order of one square foot. Included in this report is the development of a thermal imaging model, a weighted least squares thermal data smoothing algorithm, simulation and experimental flaw detection results, and an overview of the ATAC (Automated Thermal Analysis Code) software that was developed to analyze thermal inspection data

  18. Advanced Stirling Radioisotope Generator (ASRG) Thermal Power Model in MATLAB

    Science.gov (United States)

    Wang, Xiao-Yen, J.

    2012-01-01

    This paper presents a one-dimensional steady-state mathematical thermal power model of the ASRG. It aims to provide a guideline of understanding how the ASRG works and what can change its performance. The thermal dynamics and energy balance of the generator is explained using the thermal circuit of the ASRG. The Stirling convertor performance map is used to represent the convertor. How the convertor performance map is coupled in the thermal circuit is explained. The ASRG performance characteristics under i) different sink temperatures and ii) over the years of mission (YOM) are predicted using the one-dimensional model. Two Stirling converter control strategies, i) fixing the hot-end of temperature of the convertor by adjusting piston amplitude and ii) fixing the piston amplitude, were tested in the model. Numerical results show that the first control strategy can result in a higher system efficiency than the second control strategy when the ambient gets warmer or the general-purpose heat source (GPHS) fuel load decays over the YOM. The ASRG performance data presented in this paper doesn't pertain to the ASRG flight unit. Some data of the ASRG engineering unit (EU) and flight unit that are available in public domain are used in this paper for the purpose of numerical studies.

  19. 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 30...

  20. Equivalent thermal conductivity of the storage basket with spent nuclear fuel of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Alyokhina, Svitlana; Kostikov, Andriy

    2014-01-01

    Due to limitation of computation resources and/or computation time many thermal problems require to use simplified geometrical models with equivalent thermal properties. A new method for definition of equivalent thermal conductivity of spent nuclear fuel storage casks is proposed. It is based on solving the inverse heat conduction problem. For the proposed method two approaches for equivalent thermal conductivity definition were considered. In the first approach a simplified model in conjugate formulation is used, in the second approach a simplified model of solid body which allows an analytical solution is used. For safety ensuring during all time of spent nuclear fuel storage the equivalent thermal conductivity was calculated for different storage years. The calculated equivalent thermal conductivities can be used in thermal researches for dry spent nuclear fuel storage safety.

  1. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    Science.gov (United States)

    Emrich, William J., Jr.

    2014-01-01

    Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit

  2. Real-time advanced nuclear reactor core model

    International Nuclear Information System (INIS)

    Koclas, J.; Friedman, F.; Paquette, C.; Vivier, P.

    1990-01-01

    The paper describes a multi-nodal advanced nuclear reactor core model. The model is based on application of modern equivalence theory to the solution of neutron diffusion equation in real time employing the finite differences method. The use of equivalence theory allows the application of the finite differences method to cores divided into hundreds of nodes, as opposed to the much finer divisions (in the order of ten thousands of nodes) where the unmodified method is currently applied. As a result the model can be used for modelling of the core kinetics for real time full scope training simulators. Results of benchmarks, validate the basic assumptions of the model and its applicability to real-time simulation. (orig./HP)

  3. Advanced passive technology: A global standard for nuclear plant requirements

    Energy Technology Data Exchange (ETDEWEB)

    Novak, V

    1994-12-31

    Since 1984, Westinghouse has been developing AP8OO, a 800 MW, two-loop advanced passive plant, in response to an initiative established by the Electric Power Research Institute (EPRI) and the U.S. Department of Energy` (DOE). The preliminary design was cornpleved in 1989. AP6OO`s Standard Safety Analysis and Probabilistic Risk analysis Reports were submitted to the U.S. Nuclear Regulatory Commission for design certification in 1992. Design simplification is the key strategy behind the AP6OO. The basic technical concept Of simplification has resulted in a simplified reactor coolant systems, simplified plant systems, a simplified plant arrangement, reduced number of components, simplified operation and maintenance.

  4. Advanced computer graphics techniques as applied to the nuclear industry

    International Nuclear Information System (INIS)

    Thomas, J.J.; Koontz, A.S.

    1985-08-01

    Computer graphics is a rapidly advancing technological area in computer science. This is being motivated by increased hardware capability coupled with reduced hardware costs. This paper will cover six topics in computer graphics, with examples forecasting how each of these capabilities could be used in the nuclear industry. These topics are: (1) Image Realism with Surfaces and Transparency; (2) Computer Graphics Motion; (3) Graphics Resolution Issues and Examples; (4) Iconic Interaction; (5) Graphic Workstations; and (6) Data Fusion - illustrating data coming from numerous sources, for display through high dimensional, greater than 3-D, graphics. All topics will be discussed using extensive examples with slides, video tapes, and movies. Illustrations have been omitted from the paper due to the complexity of color reproduction. 11 refs., 2 figs., 3 tabs

  5. Thermal performance monitoring and assessment in Dukovany nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Madron, F. [Chemplant Technology s.r.o., Hrncirska 4, 400 01 Usti nad Labem (Czech Republic); Papuga, J. [CEZ a.s., JE Dukovany, 675 50 Dukovany (Czech Republic); Pliska, J. [I and C ENERGO a.s., Prazska 684, 674 01 Trebic (Czech Republic)

    2006-07-01

    Competition in the European electricity market forces generators to achieve - in compliance with safety and environmental standards - efficiency of production as high as possible. This efficiency or heat rate is an important indicator of both the condition of the plant equipment and the quality of plant operation. Similar thermal performance indicators can also be calculated for components of the plant equipment such as heat exchangers. However, it is not easy to quantify these indicators with sufficient precision so that the results can be used for conduct of plant operation in near-real time and for predictive maintenance. This paper describes a present state of the system monitoring and evaluating thermal performance of the reactor units in Dukovany Nuclear Power Plant. The system provides information on actual and desirable (should-be) values of thermal performance indicators for control room operators, performance engineers and maintenance planners. The system is designed to monitor steady states and has two main functions: data validation and process simulation. Data validation is based on data reconciliation methodology and carried out with Recon software by Chemplant Technology. A detailed model of the secondary side for mass and heat balancing has been made up by means of the Recon's graphical editor; now it contains roughly 300 flows and employs data of about 200 measurements. Main advantages of the data reconciliation are: - reconciled data are consistent with the model, - reconciled data are more precise than data directly measured with consequence that the thermal power of steam generators is determined with substantially lower uncertainty than before - data reconciliation represents a solid basis for detection and identification of data corrupted by gross errors. Simulation is performed with a different analytical model of plant components configured into secondary side. The model has been developed by I and C Energo. Main purposes of simulation

  6. Stability, rheology and thermal analysis of functionalized alumina- thermal oil-based nanofluids for advanced cooling systems

    International Nuclear Information System (INIS)

    Ilyas, Suhaib Umer; Pendyala, Rajashekhar; Narahari, Marneni; Susin, Lim

    2017-01-01

    Highlights: • Alumina nanoparticles are functionalized with oleic acid. • Functionalization of alumina nanoparticles gives better dispersion in thermal oil. • Thermophysical properties of nanofluids are experimentally measured. • TGA confirms the improvement in life of nanofluids. - Abstract: Thermal oils are widely used as cooling media in heat transfer processes. However, their potential has not been utilised exquisitely in many applications due to low thermal properties. Thermal oil-based nanofluids are prepared by dispersing functionalized alumina with varying concentrations of 0.5–3 wt.% to enhance thermal properties of oil for advanced cooling systems. The oleic acid coated alumina is prepared and then dispersed in the oil to overcome the aggregation of nanoparticles in base fluid. The surface characterizations of functionalized nanoparticles are performed using different analysis such as XRD, EDS, SEM, TEM and FTIR. Dispersion behaviour and agglomeration studies are conducted at natural and functionalized conditions using different analysis to ensure long-term stability of nanofluids. In addition, rheological behaviour of non-Newtonian nanofluids is studied at high shear rates (100–2000 s"−"1). Effective densities and enhancement in thermal conductivities are measured for different nanofluids concentrations. Specific heat capacity is measured using Differential Scanning Calorimetry. The correlations are developed for thermophysical properties of nanofluids. Thermogravimetric analysis is performed with respect to temperature and time to exploit the effect of the addition of nanoparticles on the degradation of nanofluids. Significant improvement in the thermal properties of oil is observed using highly stable functionalized alumina nano-additives.

  7. Affordable Development and Qualification Strategy for Nuclear Thermal Propulsion

    Science.gov (United States)

    Gerrish, Harold P., Jr.; Doughty, Glen E.; Bhattacharyya, Samit K.

    2013-01-01

    Nuclear Thermal Propulsion (NTP) is a concept which uses a nuclear reactor to heat a propellant to high temperatures without combustion and can achieve significantly greater specific impulse than chemical engines. NTP has been considered many times for human and cargo missions beyond low earth orbit. A lot of development and technical maturation of NTP components took place during the Rover/NERVA program of the 60's and early 70's. Other NTP programs and studies followed attempting to further mature the NTP concept and identify a champion customer willing to devote the funds and support the development schedule to a demonstration mission. Budgetary constraints require the use of an affordable development and qualification strategy that takes into account all the previous work performed on NTP to construct an existing database, and include lessons learned and past guidelines followed. Current guidelines and standards NASA uses for human rating chemical rocket engines is referenced. The long lead items for NTP development involve the fuel elements of the reactor and ground testing the engine system, subsystem, and components. Other considerations which greatly impact the development plans includes the National Space Policy, National Environmental Policy Act, Presidential Directive/National Security Council Memorandum #25 (Scientific or Technological Experiments with Possible Large-Scale Adverse Environmental Effects and Launch of Nuclear Systems into Space), and Safeguards and Security. Ground testing will utilize non-nuclear test capabilities to help down select components and subsystems before testing in a nuclear environment to save time and cost. Existing test facilities with minor modifications will be considered to the maximum extent practical. New facilities will be designed to meet minimum requirements. Engine and test facility requirements are based on the driving mission requirements with added factors of safety for better assurance and reliability

  8. Advanced homogenization strategies in material modeling of thermally sprayed TBCs

    International Nuclear Information System (INIS)

    Bobzin, K.; Lugscheider, E.; Nickel, R.; Kashko, T.

    2006-01-01

    Thermal barrier coatings (TBC), obtained by atmospheric plasma spraying (APS), have a complex microstructure (lamellar, porous, micro-cracked). Process parameters take an influence on this microstructure. Two methods based on the homogenization for periodic structures are presented in this article. The methods are used to calculate the effective material behavior of APS-TBCs made of partially yttria stabilized zirconia (PYSZ) depending on the microstructure. (Abstract Copyright [2006], Wiley Periodicals, Inc.)

  9. Advanced welding for closed structure. Pt. 3 The thermal approach

    Energy Technology Data Exchange (ETDEWEB)

    Sacripanti, A.; Bonanno, G.; Paoloni, M.; Sagratella, G. [ENEA Centro Ricerche Casaccia, Rome (Italy). Dipt. Innovazione; Arborino, A.; Varesi, R.; Antonucci, A. [DUNE, (Italy)

    1999-07-01

    This report describes the activities developed for the European Contract BRITE AWCS III to study the use of thermal sensing techniques to obtain an accurate detection of the internal reinforcement of the closed steel structures employed in the shipbuilding industry. After a description of the methods, normally developed in Russia, about the techniques and problems, for the thermal testing of materials in the conventional approach, a new thermal detector was utilized, a new bolometric thermo camera is introduced with a special software for the on line image analysis, there are also shown the experimental tests and results. The obtained conclusion shows that the thermal non destructive testing techniques with the new detector should be useful to assemble a complete sensing system with one ultrasonic head. [Italian] Questo rapporto descrive le attivita' sperimentali sviluppate nell'ambito del contratto europeo BRITE AWCS III, in cui si sono utilizzate tecniche termiche per ottenere un preciso rilevamento dei rinforzi interni di strutture metalliche chiuse utilizzate nell'industria delle costruzioni navali. Dopo la descrizione dei metodi sviluppati essenzialmente in Russia, circa le tecniche e i problemi riguardanti il testing termico dei materiali, e' stato introdotto un approccio innovativo basato su un nuovo sensore: una termocamera bolometrica connessa con un software dedicato per l'analisi online del setto; vengono inoltre mostrati i risultati sperimentali ottenuti. Le conclusioni ottenute mostrano che nel nuovo approccio, il testing termico non distruttivo dovrebbe essere utile per assemblare un sistema sensoriale completo che utilizzi anche un sensore di tipo ultrasonico.

  10. Advanced thermal management of diesel engines; Neues Thermomanagement beim Dieselmotor

    Energy Technology Data Exchange (ETDEWEB)

    Wenzel, Wolfgang; Becker, Michael [BorgWarner, Ludwigsburg (Germany). Konzernvorentwicklung fuer Pkw-Antriebssysteme; Shutty, John [BorgWarner, Auburn Hills (United States). Regelung und Simulation in der Konzernvorentwicklung

    2013-05-01

    The potential of thermal management with respect to CO{sub 2} reduction is given by faster warm-up of engine and drivetrain, reduced losses from water pump and fan and finally the operation of the engine in an optimal temperature range. In a new approach, BorgWarner applies a variable coolant pump and a controlled coolant valve to a conventional cooling system. Both components, as well as the viscous fan clutch, are controlled by a newly developed controls approach.

  11. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    Energy Technology Data Exchange (ETDEWEB)

    Benson, Jeff B; Marshall, Frances M [Idaho National Laboratory, Idaho Falls, ID (United States); Allen, Todd R [Univ. of Wisconsin, Madison, WI (United States)

    2012-03-15

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  12. Recent advances in nuclear medicine in endocrine oncology.

    Science.gov (United States)

    Luster, Markus; Pfestroff, Andreas; Verburg, Frederik A

    2017-01-01

    The purpose is to review recent advances concerning the role of nuclear medicine in endocrine oncology. For I therapy of thyroid cancer a thyrotropin (TSH) more than 30 mU/l has for many years been deemed a condition sine qua non. However, new data show that patients with lower TSH levels at the time of ablation have the same rate of successful ablation as those with TSH more than 30 mU/l.I-124 combined integrated positron emission tomography and computed X-ray tomography was shown to be highly accurate in predicting findings on posttherapy radioiodine scanning and was shown to have a high prognostic power.In neuroendocrine tumors, long-term complication rates of peptide receptor radionuclide therapy were reported. Furthermore first preclinical and clinical results of peptide receptor radionuclide therapy with somatostatin receptor antagonists were published.In nuclear medicine, prostate-specific membrane antigen (PSMA)-targeted radionuclide imaging and therapy is of interest. PSMA was shown to also be expressed in neoplasms of the thyroid, the adrenal glands and neuroendocrine tumors. Further individualization of thyroid cancer patient care by means of I-124-positron emission tomography and computed X-ray tomography-based selection of the therapeutic strategy is possible. I therapy might not require as intensive TSH stimulation as thought previously. For endocrine-related malignancies PSMA targeting deserves further investigation.

  13. Advanced methods in evaluation of thermal power systems effectiveness

    International Nuclear Information System (INIS)

    Barnak, N.; Jakubcek, P.; Zadrazil, J.

    1993-01-01

    The universal method for thermodynamic systems process irreversibility evaluation based on exergetic approach is elaborated in this article. The method uses the basic property of exergy as extensive state parameter -additivity. Division of the system onto some hierarchic levels is considered and relation between exergetic system characteristics and its parts is defined. There are system structure coefficients in common form expressed article they are analysed. The criteria for technical and economical optimization of the system using expressed structure coefficients are defined. In the article, there are common approaches defined for the method application in the area of nuclear power plant secondary circuits and the method is used for nuclear power plant WWER-1000 secondary circuit analysis. For this, individual exergetic characteristics of secondary circuit and its parts are expressed and some of secondary circuit parameters are optimized. Proposals for practical realisation of the results are stated in the conclusions of the article, mainly in the area of computerized evaluation of technical and economical parameters of nuclear power plant and effectiveness of its operation

  14. Studies on environment safety and application of advanced reactor for inland nuclear power plants

    International Nuclear Information System (INIS)

    Wei, L.; Jie, L.

    2014-01-01

    To study environment safety assessment of inland nuclear power plants (NPPs), the impact of environment safety under the normal operation was researched and the environment risk of serious accidents was analyzed. Moreover, the requirements and relevant provisions of site selection between international nuclear power plant and China's are comparatively studied. The conclusion was that the environment safety assessment of inland and coastal nuclear power plant have no essential difference; the advanced reactor can meet with high criteria of environment safety of inland nuclear power plants. In this way, China is safe and feasible to develop inland nuclear power plant. China's inland nuclear power plants will be as big market for advanced reactor. (author)

  15. Nuclear reactor thermal hydraulics safety analysis and thoughts on FUKUSHIMA

    International Nuclear Information System (INIS)

    Ninokata, Hisashi

    2012-01-01

    The first part of this article is to show my thoughts on the accident at Fukushima Daiichi Nuclear Power Station. It is cited from a summary of my lecture talk in Indonesia, in the beginning of the last December, 2011. This talk was based on my previous lecture and seminar talks including those delivered at MIT, June 16, at the ANS Annual Meeting in Hollywood, Florida, June 28 at NURETH-13 in Toronto, September 27, and others. The content is based on the open and latest information available to date in Japan. It may contain some erroneous or uncertain information. I tried to minimize it to my best capability. Also I tried to eliminate any critical issues or opinions that may jeopardize some people who were involved in. The latter half of this article will be excerpts of my recent R and D activities related to the safety-by-design for sodium cooled fast reactors and light water reactors, thermal hydraulics analysis focusing on the simulation-based technology, in particular subchannel analysis and computational fluid dynamics. (J.P.N.)

  16. Nuclear thermal rocket plume interactions with spacecraft. Final report

    International Nuclear Information System (INIS)

    Mauk, B.H.; Gatsonis, N.A.; Buzby, J.; Yin, X.

    1997-01-01

    This is the first study that has treated the Nuclear Thermal Rocket (NTR) effluent problem in its entirety, beginning with the reactor core, through the nozzle flow, to the plume backflow. The summary of major accomplishments is given below: (1) Determined the NTR effluents that include neutral, ionized and radioactive species, under typical NTR chamber conditions. Applied an NTR chamber chemistry model that includes conditions and used nozzle geometries and chamber conditions typical of NTR configurations. (2) Performed NTR nozzle flow simulations using a Navier-Stokes solver. We assumed frozen chemistry at the chamber conditions and used nozzle geometries and chamber conditions typical of NTR configurations. (3) Performed plume simulations using a Direct Simulation Monte Carlo (DSMC) code with chemistry. In order to account for radioactive trace species that may be important for contamination purposes we developed a multi-weighted DSMC methodology. The domain in our simulations included large regions downstream and upstream of the exit. Inputs were taken from the Navier-Stokes solutions

  17. In-place thermal annealing of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Server, W.L.

    1985-04-01

    Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

  18. Social assessment and location of nuclear and thermal power plants

    International Nuclear Information System (INIS)

    Nemoto, Kazuyasu; Nishio, Mitsuo.

    1979-01-01

    Most of the locations of nuclear and thermal power plants in Japan are depopulated villages with remote rural character, but for the development of such districts, the policy is not yet clearly established, and the appropriate measures are not taken. The living regions of residents and the production regions of enterprises are more and more estranged. Social assessment is the scientific method to perceive the future change due to the installation of power stations. The features particular to the assessment of natural environment and social environment related to the location of power stations are considered, and the technical problems involved in the method of assessment of natural environment are solved, and the actual method of assessment of social environment is developed. Then, the possibility of establishing this method and the problems in its application are investigated. The plan of developing the surroundings of power generation facilities is criticized, and the coordination of the location plan of power companies and the regional projects of municipalities is discussed. Finally, the mechanism of consensus formation concerning the location of power stations is considered, dividing into regional consensus formation and administrative consensus formation, and the possibility of instituting social assessment is examined. (Kako, I.)

  19. Advanced Computational Methods for Thermal Radiative Heat Transfer

    Energy Technology Data Exchange (ETDEWEB)

    Tencer, John; Carlberg, Kevin Thomas; Larsen, Marvin E.; Hogan, Roy E.,

    2016-10-01

    Participating media radiation (PMR) in weapon safety calculations for abnormal thermal environments are too costly to do routinely. This cost may be s ubstantially reduced by applying reduced order modeling (ROM) techniques. The application of ROM to PMR is a new and unique approach for this class of problems. This approach was investigated by the authors and shown to provide significant reductions in the computational expense associated with typical PMR simulations. Once this technology is migrated into production heat transfer analysis codes this capability will enable the routine use of PMR heat transfer in higher - fidelity simulations of weapon resp onse in fire environments.

  20. Advanced kinetics for calorimetric techniques and thermal stability screening of sulfide minerals

    International Nuclear Information System (INIS)

    Iliyas, Abduljelil; Hawboldt, Kelly; Khan, Faisal

    2010-01-01

    Thermal methods of analysis such as differential scanning calorimetry (DSC) provide a powerful methodology for the study of solid reactions. This paper proposes an improved thermal analysis methodology for thermal stability investigation of complex solid-state reactions. The proposed methodology is based on differential iso-conversional approach and involves peak separation, individual peak analysis and combination of isothermal/non-isothermal DSC measurements for kinetic analysis and prediction. The proposed thermal analysis, which coupled with Mineral Libration Analyzer (MLA) technique was employed to investigate thermal behavior of sulfide mineral oxidation. The importance of various experimental variables such as particle size, heating rate and atmosphere were investigated and discussed. The information gained from such an advanced thermal analysis method is useful for scale-up processes with potential of significant savings in plant operations, as well as in mitigating adverse environmental and safety issues arising from handling and storage of sulfide minerals.

  1. Development of expert evaluation system for advanced nuclear power stations

    International Nuclear Information System (INIS)

    Yang Mengjia; Ren Junsheng; Zhou Zhiwei; Zhang Qi; Zhu Shutang

    2005-01-01

    The wheels of the nuclear power industry in China are going to roll on the fast rails soon: It is important to establish an expert evaluation system for optimally selecting new types of NPPs in the future based on scientific evaluation. Under such a background, an expert evaluation system is developed to evaluate Various types of NPPs for the following two decades by taking into account the advices of selected top specialists in Chinese nuclear power industry, and by adopting different evaluation methods. The evaluation system is designed to use the database-MS-SQL Server or Oracle for manipulating experts' knowledge and objective data for scaling NPPs by user-selected evaluation methods for evaluating NPPs. The system includes two subsystems: one is called the Data Management Module and the other called the Evaluation Arithmetic Module. The Data Management Module is used to manage expert knowledge in forms of item marks of the NPPs given by experts of nuclear engineering. The Evaluation Arithmetic Module is used to analyze and deal with the evaluation based on the data in the Database by certain designed evaluation method, and to provide the evaluation results. Based on the users' selection, the evaluation of the NPPs is processed, and the results can be dynamically listed in table or displayed in graphics. So it can be used as a consulting tool to help users make optimal selection of prospective NPPs. The system is developed in architecture of the mixture framework C/S(Client/Server) and B/S(Browse/Server), thus is safe, advanced, extensible and user- friendly. The present system is computer network based, so it can be used not only in personal PC and local network, but also in the Web Server which can be browsed by remote users. To develop such an evaluation system is an initiative R and D project in Chinese nuclear R and D activities. Although the evaluation system is far from perfect, however, the preliminary evaluation results produced by the evaluation

  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 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).

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

    International Nuclear Information System (INIS)

    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

    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. It has been looking for these techniques to expand the working area of human, raise the working efficiencies of remote task to the highest degree, and enhance the industrial

  4. Advanced nuclear systems in comparison; Fortgeschrittene nukleare Systeme im Vergleich

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R.; Foskolos, K.; Goetzmann, C.; Kroeger, W.; Stanculescu, A.; Wydler, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-09-01

    This study aims at a comparison of future reactor concepts, paying particular attention to aspects of safety, of the fuel cycle, the economics, the experience-base and the state of development. Representative examples of typical development lines, that could possibly be `of interest` within a time horizon of 50 years were selected for comparison. This can be divided into three phases: - Phase I includes the next 10 years and will be characterised mainly by evolutionary developments of light water reactors (LWR) of large size; representative: EPR, - Phase II: i.e. the time between 2005 and 2020 approximately, encompasses the forecasted doubling of today`s world-wide installed nuclear capacity; along with evolutionary reactors, innovative systems like AP600, PIUS, MHTGR, EFR will emerge, - Phase III covers the time between 2020 and 2050 and is characterised by the issue of sufficient fissile material resources; novel fast reactor systems including hybrid systems can, thus, become available; representatives: IFR, EA, ITER (the latter being). The evaluated concepts foresee partly different fuel cycles. Fission reactors can be operated in principle on the basis of either a Uranium-Plutonium-cycle or a Thorium-Uranium-cycle, while combinations of these cycles among them or with other reactor concepts than proposed are possible. With today`s nuclear park (comprising mainly LWRs), the world-wide plutonium excess increases annually by about 100 t. Besides strategies based on reprocessing like: - recycling in thermal and fast reactors with mixed oxide fuels, - plutonium `burning` in reactors with novel fuels without uranium or in `hybrid` systems, allowing a reduction of this excess, direct disposal of spent fuel elements including their plutonium content (`one-through`) is being considered. (author) figs., tabs., 32 refs.

  5. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    International Nuclear Information System (INIS)

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

  6. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [Brookhaven National Lab., Upton, NY (United States)] [comp.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  7. Innovative nuclear thermal rocket concept utilizing LEU fuel for space application

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Venneri, Paolo; Choi, Jae Young; Jeong, Yong Hoon; Chang, Soon Heung

    2015-01-01

    Space is one of the best places for humanity to turn to keep learning and exploiting. A Nuclear Thermal Rocket (NTR) is a viable and more efficient option for human space exploration than the existing Chemical Rockets (CRs) which are highly inefficient for long-term manned missions such as to Mars and its satellites. NERVA derived NTR engines have been studied for the human missions as a mainstream in the United States of America (USA). Actually, the NERVA technology has already been developed and successfully tested since 1950s. The state-of-the-art technology is based on a Hydrogen gas (H_2) cooled high temperature reactor with solid core utilizing High-Enriched Uranium (HEU) fuel to reduce heavy metal mass and to use fast or epithermal neutron spectrums enabling simple core designs. However, even though the NTR designs utilizing HEU is the best option in terms of rocket performance, they inevitably provoke nuclear proliferation obstacles on all Research and Development (R and D) activities by civilians and non-nuclear weapon states, and its eventual commercialization. To surmount the security issue to use HEU fuel for a NTR, a concept of the innovative NTR engine, Korea Advanced NUclear Thermal Engine Rocket utilizing Low-Enriched Uranium fuel (KANUTER-LEU) is presented in this paper. The design goal of KANUTER-LEU is to make use of a LEU fuel for its compact reactor, but does not sacrifice the rocket performance relative to the traditional NTRs utilizing HEU. KANUTER-LEU mainly consists of a fission reactor utilizing H_2 propellant, a propulsion system and an optional Electricity Generation System as a bimodal engine. To implement LEU fuel for the reactor, the innovative engine adopts W-UO_2 CERMET fuel to drastically increase uranium density and thermal neutron spectrum to improve neutron economy in the core. The moderator and structural material selections also consider neutronic and thermo-physical characteristics to reduce non-fission neutron loss and

  8. Safety related terms for advanced nuclear plants; Terminos relacionados con la seguridad para centrales nucleares avanzadas

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-01

    The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety.

  9. A study on the advanced statistical core thermal design methodology

    International Nuclear Information System (INIS)

    Lee, Seung Hyuk

    1992-02-01

    A statistical core thermal design methodology for generating the limit DNBR and the nominal DNBR is proposed and used in assessing the best-estimate thermal margin in a reactor core. Firstly, the Latin Hypercube Sampling Method instead of the conventional Experimental Design Technique is utilized as an input sampling method for a regression analysis to evaluate its sampling efficiency. Secondly and as a main topic, the Modified Latin Hypercube Sampling and the Hypothesis Test Statistics method is proposed as a substitute for the current statistical core thermal design method. This new methodology adopts 'a Modified Latin Hypercube Sampling Method' which uses the mean values of each interval of input variables instead of random values to avoid the extreme cases that arise in the tail areas of some parameters. Next, the independence between the input variables is verified through 'Correlation Coefficient Test' for statistical treatment of their uncertainties. And the distribution type of DNBR response is determined though 'Goodness of Fit Test'. Finally, the limit DNBR with one-sided 95% probability and 95% confidence level, DNBR 95/95 ' is estimated. The advantage of this methodology over the conventional statistical method using Response Surface and Monte Carlo simulation technique lies in its simplicity of the analysis procedure, while maintaining the same level of confidence in the limit DNBR result. This methodology is applied to the two cases of DNBR margin calculation. The first case is the application to the determination of the limit DNBR where the DNBR margin is determined by the difference between the nominal DNBR and the limit DNBR. The second case is the application to the determination of the nominal DNBR where the DNBR margin is determined by the difference between the lower limit value of the nominal DNBR and the CHF correlation limit being used. From this study, it is deduced that the proposed methodology gives a good agreement in the DNBR results

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

    International Nuclear Information System (INIS)

    Lee, Jong Min; Lee, Yong Bum; Kim, Woong Ki; Park, Soon Yong; Kim, Seung Ho; Kim, Chang Hoi; Hwang, Suk Yeoung; Kim, Byung Soo; Lee, Young Kwang

    1994-07-01

    In this year (the second year of this project), researches and development have been carried out to establish the essential key technologies applied to robot system for nuclear industry. In the area of robot vision, in order to construct stereo vision system necessary to tele-operation, stereo image acquisition camera module and stereo image displayer have been developed. Stereo matching and storing programs have been developed to analyse stereo images. According to the result of tele-operation experiment, operation efficiency has been enhanced about 20% by using the stereo vision system. In a part of object recognition, a tele-operated robot system has been constructed to evaluate the performance of the stereo vision system and to develop the vision algorithm to automate nozzle dam operation. A nuclear fuel rod character recognition system has been developed by using neural network. As a result of perfomance evaluation of the recognition system, 99% recognition rate has been achieved. In the area of sensing and intelligent control, temperature distribution has been measured by using the analysis of thermal image histogram and the inspection algorithm has been developed to determine of the state be normal or abnormal, and the fuzzy controller has been developed to control the compact mobile robot designed for path moving on block-typed path. (Author)

  11. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  12. Approaches to passive safety in advanced thermal reactors

    International Nuclear Information System (INIS)

    Moses, D.L.

    1986-01-01

    Since 1980, there has been a proliferation of thermal reactor designs which incorporate passive safety features. The evolution of this trend is briefly traced, and the nature of various passive safety features is discussed with regard to how they have been incorporated into evolving design concepts. The key aspects of the passive safety features include reduced core power density, enhanced passive heat sinks, inherent assured shutdown mechanisms, elimination/minimization of potential leak paths from the primary coolant systems, enhanced robustness of fuel elements and improved coolant chemistry and component materials. An increased reliance on purely passive safety features typically translates into larger reactor structures at reduced power ratings. Proponents of the most innovative concepts seek to offset the increased costs by simplifying licensing requirements and reducing construction time

  13. Advanced I and C systems for nuclear power plants feedback of experience

    International Nuclear Information System (INIS)

    Prehler, H.J.

    2001-01-01

    Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally specific requirements arising from nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS.(author)

  14. Advanced energy system with nuclear reactors as an energy source

    International Nuclear Information System (INIS)

    Kato, Y.; Ishizuka, T.; Nikitin, K.

    2007-01-01

    recovery system is also applicable to a fast reactor (FR) with a supercritical CO 2 gas turbine that achieves higher cycle efficiency than conventional sodium cooled FRs with steam turbines. The FR will eliminate problems of conventional FRs related to safety, plant maintenance, and construction costs. The FR consumes efficiently trans-uranium elements (TRU) produced in light water reactors as fuel and reduce long-lived radioactive wastes or environmental loads of long term geological disposal. An Advanced Energy System (AES) with nuclear reactors as an energy source has been proposed which supply electricity and heat to cities. The AES has three objectives: 1. Save energy resources and reduce green house gas emissions, attaining total energy utilization efficiency higher than 85% through waste heat recovery and utilization. 2. Foster a recycling society that produces methane and methanol for fuel cells from waste products of cities and farms. 3. Consume TRU produced in LWRs as fuel for FRs, and reduce long-lived radioactive wastes or environmental loads of long term geological disposal. References 1. Y. Kato, T. Nitawaki and K. Fujima, 'Zero Waste Heat Release Nuclear Cogeneration System, 'Proc. 2003 Intl. Congress on Advanced Nuclear Power Plants (ICAPP'03), Cordoba, Spain, May 4-7, 2003, Paper 3313. 2. Y. Kato, T. Nitawaki and Y. Muto, 'Medium Temperature Carbon Dioxide Gas Turbine Reactor, 'Nucl. Eng. Design, 230, pp. 195-207 (2004). 3. H. N. Tran and Y. Kato, 'New 2 37Np Burning Strategy in a Supercritical CO 2 Cooled Fast Reactor Core Attaining Zero Burnup Reactivity Loss,' Proc. American Nuclear Society's Topical Meeting on Reactor Physics (PHYSOR 2006), Vancouver, British Columbia, Canada, September 10-14, 2006

  15. Advanced Reactor Technology Options for Utilization and Transmutation of Actinides in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    2009-09-01

    Renewed interest in the potential of nuclear energy to contribute to a sustainable worldwide energy mix is strengthening the IAEA's statutory role in fostering the peaceful uses of nuclear energy, in particular the need for effective exchanges of information and collaborative research and technology development among Member States on advanced nuclear power technologies (Articles III-A.1 and III-A.3). The major challenges facing the long term development of nuclear energy as a part of the world's energy mix are improvement of the economic competitiveness, meeting increasingly stringent safety requirements, adhering to the criteria of sustainable development, and public acceptability. The concern linked to the long life of many of the radioisotopes generated from fission has led to increased R and D efforts to develop a technology aimed at reducing the amount of long lived radioactive waste through transmutation in fission reactors or accelerator driven hybrids. In recent years, in various countries and at an international level, more and more studies have been carried out on advanced and innovative waste management strategies (i.e. actinide separation and elimination). Within the framework of the Project on Technology Advances in Fast Reactors and Accelerator Driven Systems (http://www.iaea.org/inisnkm/nkm/aws/fnss/index.html), the IAEA initiated a number of activities on utilization of plutonium and transmutation of long lived radioactive waste, accelerator driven systems, thorium fuel options, innovative nuclear reactors and fuel cycles, non-conventional nuclear energy systems, and fusion/fission hybrids. These activities are implemented under the guidance and with the support of the IAEA Nuclear Energy Department's Technical Working Group on Fast Reactors (TWG-FR). This publication compiles the analyses and findings of the Coordinated Research Project (CRP) on Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste (2002

  16. Advanced high-temperature thermal energy storage media for industrial applications

    Science.gov (United States)

    Claar, T. D.; Waibel, R. T.

    1982-02-01

    An advanced thermal energy storage media concept based on use of carbonate salt/ceramic composite materials is being developed for industrial process and reject heat applications. The composite latent/sensible media concept and its potential advantages over state of the art latent heat systems is described. Media stability requirements, on-going materials development efforts, and planned thermal energy storage (TES) performance evaluation tests are discussed.

  17. High Thermal Conductivity NARloy-Z-Diamond Composite Combustion Chamber Liner For Advanced Rocket Engines

    Science.gov (United States)

    Bhat, Biliyar N.; Ellis, David; Singh, Jogender

    2014-01-01

    Advanced high thermal conductivity materials research conducted at NASA Marshall Space Flight Center (MSFC) with state of the art combustion chamber liner material NARloy-Z showed that its thermal conductivity can be increased significantly by adding diamond particles and sintering it at high temperatures. For instance, NARloy-Z containing 40 vol. percent diamond particles, sintered at 975C to full density by using the Field assisted Sintering Technology (FAST) showed 69 percent higher thermal conductivity than baseline NARloy-Z. Furthermore, NARloy-Z-40vol. percent D is 30 percent lighter than NARloy-Z and hence the density normalized thermal conductivity is 140 percent better. These attributes will improve the performance and life of the advanced rocket engines significantly. By one estimate, increased thermal conductivity will directly translate into increased turbopump power up to 2X and increased chamber pressure for improved thrust and ISP, resulting in an expected 20 percent improvement in engine performance. Follow on research is now being conducted to demonstrate the benefits of this high thermal conductivity NARloy-Z-D composite for combustion chamber liner applications in advanced rocket engines. The work consists of a) Optimizing the chemistry and heat treatment for NARloy-Z-D composite, b) Developing design properties (thermal and mechanical) for the optimized NARloy-Z-D, c) Fabrication of net shape subscale combustion chamber liner, and d) Hot fire testing of the liner for performance. FAST is used for consolidating and sintering NARlo-Z-D. The subscale cylindrical liner with built in channels for coolant flow is also fabricated near net shape using the FAST process. The liner will be assembled into a test rig and hot fire tested in the MSFC test facility to determine performance. This paper describes the development of this novel high thermal conductivity NARloy-Z-D composite material, and the advanced net shape technology to fabricate the combustion

  18. THEHYCO-3DT: Thermal hydrodynamic code for the 3 dimensional transient calculation of advanced LMFBR core

    Energy Technology Data Exchange (ETDEWEB)

    Vitruk, S.G.; Korsun, A.S. [Moscow Engineering Physics Institute (Russian Federation); Ushakov, P.A. [Institute of Physics and Power Engineering, Obninsk (R)] [and others

    1995-09-01

    The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.

  19. THEHYCO-3DT: Thermal hydrodynamic code for the 3 dimensional transient calculation of advanced LMFBR core

    International Nuclear Information System (INIS)

    Vitruk, S.G.; Korsun, A.S.; Ushakov, P.A.

    1995-01-01

    The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors

  20. Nuclear grade cable thermal life model by time temperature superposition algorithm based on Matlab GUI

    International Nuclear Information System (INIS)

    Lu Yanyun; Gu Shenjie; Lou Tianyang

    2014-01-01

    Background: As nuclear grade cable must endure harsh environment within design life, it is critical to predict cable thermal life accurately owing to thermal aging, which is one of dominant factors of aging mechanism. Purpose: Using time temperature superposition (TTS) method, the aim is to construct nuclear grade cable thermal life model, predict cable residual life and develop life model interactive interface under Matlab GUI. Methods: According to TTS, nuclear grade cable thermal life model can be constructed by shifting data groups at various temperatures to preset reference temperature with translation factor which is determined by non linear programming optimization. Interactive interface of cable thermal life model developed under Matlab GUI consists of superposition mode and standard mode which include features such as optimization of translation factor, calculation of activation energy, construction of thermal aging curve and analysis of aging mechanism., Results: With calculation result comparison between superposition and standard method, the result with TTS has better accuracy than that with standard method. Furthermore, confidence level of nuclear grade cable thermal life with TTS is higher than that with standard method. Conclusion: The results show that TTS methodology is applicable to thermal life prediction of nuclear grade cable. Interactive Interface under Matlab GUI achieves anticipated functionalities. (authors)

  1. Framatome advanced nuclear power-benefits for our clients from the new company

    International Nuclear Information System (INIS)

    Weber, P.

    2001-01-01

    Framatome ANP (Advanced Nuclear Power) merges the complementary strengths of two global nuclear industry leaders Framatome and Siemens - offering clients the best technological solutions for safe, reliable and economical plant performance. With a combined workforce of 13,300 skilled individuals, Framatome ANP is now the nuclear industry's leading supplier. Serving as Original Equipment Manufacturer (OEM) for more than 90 reactors that provide about 30% of the world's total installed nuclear power capacity, our experienced resources remain focused on the local needs of individual clients, wherever in the world they may be. The Company main business used to be turnkey construction of complete Nuclear Power plants, BWR and PWR capabilities, heavy equipment manufacturing, comprehensive I and C capabilities, and also expertise and knowledge of VVER. Framatome ANP will benefit in all of its fields of activity of the experience gained through Framatome and Siemens' collaboration on the next generation reactor, the EPR, as well as on steam generators replacements and or modernization of VVER. Framatome ANP nuclear fuel designs for both PWR and BWR plants provide innovative features and world-leading performance. Framatome ANP is organized according a matrix organization with: - 4 Business Groups (Project and Engineering, Service, Nuclear Fuel, Mechanical Equipment) - 3 Regional Divisions (Framatome Advanced Nuclear Power S.A.S., France; Framatome Advanced Nuclear Power GmbH, Germany; Framatome Advanced Nuclear Power Inc., USA). By 30th January 2001 Siemens Nuclear Power GmbH, founded in 2000 as successor of the Nuclear Division of Siemens Power Generation Group (KWU), was renamed to Framatome Advanced Nuclear Power GmbH forming the German part of the world wide acting company. Over the past 40 years 23 nuclear power plants all around the world - not only pressurized and boiling water reactors, but also two heavy-watermoderated reactors have been designed, constructed and

  2. Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

    2009-01-01

    R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

  3. Advanced Thermal Storage for Central Receivers with Supercritical Coolants

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Bruce D.

    2010-06-15

    The principal objective of the study is to determine if supercritical heat transport fluids in a central receiver power plant, in combination with ceramic thermocline storage systems, offer a reduction in levelized energy cost over a baseline nitrate salt concept. The baseline concept uses a nitrate salt receiver, two-tank (hot and cold) nitrate salt thermal storage, and a subcritical Rankine cycle. A total of 6 plant designs were analyzed, as follows: Plant Designation Receiver Fluid Thermal Storage Rankine Cycle Subcritical nitrate salt Nitrate salt Two tank nitrate salt Subcritical Supercritical nitrate salt Nitrate salt Two tank nitrate salt Supercritical Low temperature H2O Supercritical H2O Two tank nitrate salt Supercritical High temperature H2O Supercritical H2O Packed bed thermocline Supercritical Low temperature CO2 Supercritical CO2 Two tank nitrate salt Supercritical High temperature CO2 Supercritical CO2 Packed bed thermocline Supercritical Several conclusions have been drawn from the results of the study, as follows: 1) The use of supercritical H2O as the heat transport fluid in a packed bed thermocline is likely not a practical approach. The specific heat of the fluid is a strong function of the temperatures at values near 400 °C, and the temperature profile in the bed during a charging cycle is markedly different than the profile during a discharging cycle. 2) The use of supercritical CO2 as the heat transport fluid in a packed bed thermocline is judged to be technically feasible. Nonetheless, the high operating pressures for the supercritical fluid require the use of pressure vessels to contain the storage inventory. The unit cost of the two-tank nitrate salt system is approximately $24/kWht, while the unit cost of the high pressure thermocline system is nominally 10 times as high. 3) For the supercritical fluids, the outer crown temperatures of the receiver tubes are in the range of 700 to 800 °C. At temperatures of 700 °C and above

  4. Advanced methods of quality control in nuclear fuel fabrication

    International Nuclear Information System (INIS)

    Onoufriev, Vladimir

    2004-01-01

    Under pressure of current economic and electricity market situation utilities implement more demanding fuel utilization schemes including higher burn ups and thermal rates, longer fuel cycles and usage of Mo fuel. Therefore, fuel vendors have recently initiated new R and D programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel. In the beginning of commercial fuel fabrication, emphasis was given to advancements in Quality Control/Quality Assurance related mainly to product itself. During recent years, emphasis was transferred to improvements in process control and to implementation of overall Total Quality Management (TQM) programmes. In the area of fuel quality control, statistical control methods are now widely implemented replacing 100% inspection. This evolution, some practical examples and IAEA activities are described in the paper. The paper presents major findings of the latest IAEA Technical Meetings (TMs) and training courses in the area with emphasis on information received at the TM and training course held in 1999 and other latest publications to provide an overview of new developments in process/quality control, their implementation and results obtained including new approaches to QC

  5. 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

  6. Ceramic thermal wind sensor based on advanced direct chip attaching package

    International Nuclear Information System (INIS)

    Zhou Lin; Qin Ming; Chen Shengqi; Chen Bei

    2014-01-01

    An advanced direct chip attaching packaged two-dimensional ceramic thermal wind sensor is studied. The thermal wind sensor chip is fabricated by metal lift-off processes on the ceramic substrate. An advanced direct chip attaching (DCA) packaging is adopted and this new packaged method simplifies the processes of packaging further. Simulations of the advanced DCA packaged sensor based on computational fluid dynamics (CFD) model show the sensor can detect wind speed and direction effectively. The wind tunnel testing results show the advanced DCA packaged sensor can detect the wind direction from 0° to 360° and wind speed from 0 to 20 m/s with the error less than 0.5 m/s. The nonlinear fitting based least square method in Matlab is used to analyze the performance of the sensor. (semiconductor devices)

  7. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Benson, J.B.; Foster, J.A.; Marshall, F.M.; Meyer, M.K.; Thelen, M.C.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  8. Generic Repository Concepts and Thermal Analysis for Advanced Fuel Cycles - 12477

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Laboratories, P.O. Box 5800 MS 0736, Albuquerque, NM 87185 (United States); Blink, James [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551-0808 (United States); Carter, Joe [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States); Fratoni, Massimiliano; Greenberg, Harris; Sutton, Mark [Lawrence Livermore National Laboratory (United States); Howard, Robert [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

    2012-07-01

    A geologic disposal concept for spent nuclear fuel (SNF) or high-level waste (HLW) consists of three components: waste inventory, geologic setting, and concept of operations. A set of reference geologic disposal concepts has been developed by the U.S. Department of Energy (DOE), Used Fuel Disposition campaign. Reference concepts are identified for crystalline rock, clay/shale, bedded salt, and deep borehole (crystalline basement) geologic settings. These were analyzed for waste inventory cases representing a range of waste types that could be produced by advanced nuclear fuel cycles. Concepts of operation consisting of emplacement mode, repository layout, and engineered barrier descriptions, were selected based on international progress. All of these disposal concepts are enclosed emplacement modes, whereby waste packages are in direct contact with encapsulating engineered or natural materials. Enclosed modes have less capacity to dissipate heat than open modes such as that proposed for a repository at Yucca Mountain. Thermal analysis has identified important relationships between waste package size and capacity, and the duration of surface decay storage needed to meet temperature limits for different disposal concepts. For the crystalline rock and clay/shale repository concepts, a waste package surface temperature limit of 100 deg. C was assumed to prevent changes in clay-based buffer material or clay-rich host rock. Surface decay storage of 50 to 100 years is needed for disposal of high-burnup LWR SNF in 4-PWR packages, or disposal of HLW glass from reprocessing LWR uranium oxide (UOX) fuel. High-level waste (HLW) from reprocessing of metal fuel used in a fast reactor could be disposed after decay storage of 50 years or less. For disposal in salt the rock thermal conductivity is significantly greater, and higher temperatures (200 deg. C) can be tolerated at the waste package surface. Decay storage of 10 years or less is needed for high-burnup LWR SNF in 4-PWR

  9. Structure design and realization of advanced nuclear reactor expert evaluation system

    International Nuclear Information System (INIS)

    Gao Bin; Zhou Zhiwei; Gu Junyang

    2007-01-01

    Advanced nuclear reactor expert evaluation system is the initial practice of software on nuclear power plants evaluation system. The system was developed in C++ code under the Visual Studio Net environment, and it used Model-View-Control (MVC) pattern as its basic frame. The system was used to access the advanced nuclear reactor in China. Available results illustrate that the frame of the system is feasible and effective. (authors)

  10. Several aspects of the effect of nuclear power engineering and thermal power engineering on the environment

    Energy Technology Data Exchange (ETDEWEB)

    Malenchenko, A F

    1979-01-01

    A survey is made of the comparative effect of nuclear power engineering and thermal power engineering on environment and man. The most significant approaches to solution of radio-ecological problems of APS are found.

  11. Fabrication and Testing of Nuclear-Thermal Propulsion Ground Test Hardware, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Efficient nuclear-thermal propulsion requires heating a low molecular weight gas, typically hydrogen, to high temperature and expelling it through a nozzle. The...

  12. Hydrogen Wave Heater for Nuclear Thermal Propulsion Component Testing, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA has identified Nuclear Thermal Propulsion (NTP) as an approach that can provide the fastest trip times to Mars and as the preferred concept for human space...

  13. Hydrogen Wave Heater for Nuclear Thermal Propulsion Component Testing, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA has identified Nuclear Thermal Propulsion (NTP) as a propulsion concept which could provide the fastest trip times to Mars and as the preferred concept for...

  14. Improved CVD Coatings for Carbide Based Nuclear Thermal Propulsion Fuel Elements, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — One of the great hurdles to further development and evaluation of nuclear thermal propulsion systems is the issue surrounding the release of radioactive material...

  15. Extreme Temperature Radiation Tolerant Instrumentation for Nuclear Thermal Propulsion Engines, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of this proposal is to develop and commercialize a high reliability, high temperature smart neutron flux sensor for NASA Nuclear Thermal Propulsion...

  16. Development of Processing Techniques for Advanced Thermal Protection Materials

    Science.gov (United States)

    Selvaduray, Guna; Cox, Michael; Srinivasan, Vijayakumar

    1997-01-01

    Thermal Protection Materials Branch (TPMB) has been involved in various research programs to improve the properties and structural integrity of the existing aerospace high temperature materials. Specimens from various research programs were brought into the analytical laboratory for the purpose of obtaining and refining the material characterization. The analytical laboratory in TPMB has many different instruments which were utilized to determine the physical and chemical characteristics of materials. Some of the instruments that were utilized by the SJSU students are: Scanning Electron Microscopy (SEM), Energy Dispersive X-ray analysis (EDX), X-ray Diffraction Spectrometer (XRD), Fourier Transform-Infrared Spectroscopy (FTIR), Ultra Violet Spectroscopy/Visible Spectroscopy (UV/VIS), Particle Size Analyzer (PSA), and Inductively Coupled Plasma Atomic Emission Spectrometer (ICP-AES). The above mentioned analytical instruments were utilized in the material characterization process of the specimens from research programs such as: aerogel ceramics (I) and (II), X-33 Blankets, ARC-Jet specimens, QUICFIX specimens and gas permeability of lightweight ceramic ablators. In addition to analytical instruments in the analytical laboratory at TPMB, there are several on-going experiments. One particular experiment allows the measurement of permeability of ceramic ablators. From these measurements, physical characteristics of the ceramic ablators can be derived.

  17. Thermal-hydraulic Experiments for Advanced Physical Model Development

    International Nuclear Information System (INIS)

    Song, Chul Hwa; Baek, W. P.; Yoon, B. J.

    2010-04-01

    The improvement of prediction models is needed to enhance the safety analysis capability through the fine measurements of local phenomena. To improve the two-phase interfacial area transport model, the various experiments were carried out used SUBO and DOBO. 2x2 and 6x6 rod bundle test facilities were used for the experiment on the droplet behavior. The experiments on the droplet behavior inside a heated rod bundle were focused on the break-up of droplets induced by a spacer grid in a rod bundle geometry. The experiments used GIRLS and JICO and CFD analysis were carried out to comprehend the local condensation of steam jet, turbulent jet induced by condensation and the thermal mixing in a pool. An experimental database of the CHF (Critical Heat Flux) and PDO (Post-dryout) had been constructed. The mechanism of the heat transfer enhancement by surface modifications in nano-fluid was investigated in boiling mode and rapid quenching mode. The special measurement techniques were developed. They are Double -sensor optical void probe, Optic Rod, PIV technique and UBIM system

  18. Thermal-hydraulic Experiments for Advanced Physical Model Development

    International Nuclear Information System (INIS)

    Song, Chulhwa

    2012-04-01

    The improvement of prediction models is needed to enhance the safety analysis capability through experimental database of local phenomena. To improve the two-phase interfacial area transport model, the various experiments were carried out with local two-phase interfacial structure test facilities. 2 Χ 2 and 6 Χ 6 rod bundle test facilities were used for the experiment on the droplet behavior. The experiments on the droplet behavior inside a heated rod bundle geometry. The experiments used GIRLS and JICO and CFD analysis were carried out to comprehend the local condensation of steam jet, turbulent jet induced by condensation and the thermal mixing in a pool. In order to develop a model for key phenomena of newly adapted safety system, experiments for boiling inside a pool and condensation in horizontal channel have been performed. An experimental database of the CHF (Critical Heat Flux) and PDO (Post-dryout) was constructed. The mechanism of the heat transfer enhancement by surface modifications in nano-fluid was investigated in boiling mode and rapid quenching mode. The special measurement techniques were developed. They are Double-sensor optical void probe, Optic Rod, PIV technique and UBIM system

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

    International Nuclear Information System (INIS)

    Labib, Satira; King, Jeffrey

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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