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Sample records for advanced gas cooled

  1. Advanced gas-cooled reactors (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

    1981-01-15

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

  2. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    Keen, Barry A.

    1990-01-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  3. Advanced gas cooled reactors - Designing for safety

    Energy Technology Data Exchange (ETDEWEB)

    Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

    1990-07-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

  4. Gas-cooled reactors for advanced terrestrial applications

    International Nuclear Information System (INIS)

    Kesavan, K.; Lance, J.R.; Jones, A.R.; Spurrier, F.R.; Peoples, J.A.; Porter, C.A.; Bresnahan, J.D.

    1986-01-01

    Conceptual design of a power plant on an inert gas cooled nuclear coupled to an open, air Brayton power conversion cycle is presented. The power system, called the Westinghouse GCR/ATA (Gas-Cooled Reactors for Advanced Terrestrial Applications), is designed to meet modern military needs, and offers the advantages of secure, reliable and safe electrical power. The GCR/ATA concept is adaptable over a range of 1 to 10 MWe power output. Design descriptions of a compact, air-transportable forward base unit for 1 to 3 MWe output and a fixed-base, permanent installation for 3 to 10 MWe output are presented

  5. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-12-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  6. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-06-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  7. 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

  8. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  9. The development of advanced gas cooled reactor iodine adsorber systems

    International Nuclear Information System (INIS)

    Meddings, P.

    1986-01-01

    Advanced Gas Cooled Reactors (AGRs) are provided with plants to process the carbon dioxide coolant prior to its discharge to atmosphere. Included in these are beds of granular activated charcoal, contained within a suitable pressure vessel, through which the high pressure carbon dioxide is passed for the purpose of retaining iodine and iodine-containing compounds. Carry-over carbon dust from the adsorption beds was identified during active in-situ commissioning testing, radio-iodine being transported with the particulate material due to gross disturbance of the adsorber carbon bed and displacement of the vessel internals. The methods used to identify the causes of the problems and find solutions are described. A development programme for the Heysham-2 and Torness reactors iodine adsorber units was set up to identify a method of de-dusting granular charcoal and develop it for full-scale use, of assess the effect under conditions of high gas density of approach velocity on charcoal fines production and to establish the pressure drop characteristics of a packed granular bed and to develop an effective design of inlet gas diffuser manifold to ensure an acceptable velocity distribution. This has involved the construction of a small scale high pressure carbon dioxide rig and development of an air flow model. This work is described. (UK)

  10. Description of the advanced gas cooled type of reactor (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

  11. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    Nonboel, E.

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  12. Measurement of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Sandalls, F.J.

    1978-03-01

    Sulphur is an important element in some food chains and the release of radioactive sulphur to the environment must be closely controlled if the chemical form is such that it is available or potentially available for entering food chains. The presence of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor warranted a study to assess the quantity and chemical form of the radioactive sulphur in order to estimate the magnitude of the potential environmental hazard which might arise from the release of coolant gas from Civil Advanced Gas-Cooled Reactors. A combination of gas chromatographic and radiochemical analyses revealed carbonyl sulphide to be the only sulphur-35 compound present in the coolant gas of the Windscale Reactor. The concentration of carbonyl sulphide was found to lie in the range 40 to 100 x 10 -9 parts by volume and the sulphur-35 specific activity was about 20 mCi per gramme. The analytical techniques are described in detail. The sulphur-35 appears to be derived from the sulphur and chlorine impurities in the graphite. A method for the preparation of carbonyl sulphide labelled with sulphur-35 is described. (author)

  13. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

    2013-01-01

    Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

  14. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Vakilian, M.

    1977-05-01

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  15. Methods for studying fuel management in advanced gas cooled reactors

    International Nuclear Information System (INIS)

    Buckler, A.N.; Griggs, C.F.; Tyror, J.G.

    1971-07-01

    The methods used for studying fuel and absorber management problems in AGRs are described. The basis of the method is the use of ARGOSY lattice data in reactor calculations performed at successive time steps. These reactor calculations may be quite crude but for advanced design calculations a detailed channel-by-channel representation of the whole core is required. The main emphasis of the paper is in describing such an advanced approach - the ODYSSEUS-6 code. This code evaluates reactor power distributions as a function of time and uses the information to select refuelling moves and determine controller positions. (author)

  16. Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

    1982-01-01

    Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

  17. Seismic snubber reduction on advanced gas-cooled reactor pipework

    International Nuclear Information System (INIS)

    Kennedy, P.A.; Harkin, N.J.

    1989-01-01

    Recent advances in pipework dynamic analysis procedures have enabled a more realistic approach to be taken to the design of pipework under earthquake loadings. In particular, it is proving possible to reduce the number of seismic snubbers employed to limit pipework displacements. This paper presents the background to, and outcome of, a snubber optimisation study performed for the main steam pipework system at Torness Nuclear Power Station. (author)

  18. Optimization of advanced gas-cooled reactor fuel performance by a stochastic method

    International Nuclear Information System (INIS)

    Parks, G.T.

    1987-01-01

    A brief description is presented of a model representing the in-core behaviour of a single advanced gas-cooled reactor fuel channel, developed specifically for optimization studies. The performances of the only suitable Numerical Algorithms Group (NAG) library package and a Metropolis algorithm routine on this problem are discussed and contrasted. It is concluded that, for the problem in question, the stochastic Metropolis algorithm has distinct advantages over the deterministic NAG routine. (author)

  19. Effect of horizontal flow on the cooling of the moderator brick in the advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    Ganesan, P.; He, S.; Hamad, F.; Gotts, J.

    2011-01-01

    The paper reports an investigation of the effect of the horizontal cross flow on the temperature of the moderator brick in UK Advanced Gas-cooled Reactor (AGR) using computational fluid dynamics (CFD) with a conjugate heat transfer model for the solid and fluid. The commercial software package of ANSYS Fluent is used for this purpose. The CFD model comprises the full axial length of one-half of a typical fuel channel (assuming symmetry) and part of neighbouring channels on either side. Two sets of simulations have been carried out, namely, one with cross flow and one without cross flow. The effect of cross flow has subsequently been derived by comparing the results from the two groups of simulations. The study shows that a small cross flow can have a significant effect on the cooling of the graphite brick, causing the peak temperature of the brick to reduce significantly. Two mechanisms are identified to be responsible for this. Firstly, the small cross flow causes a significant redistribution of the main axial downward flow and this leads to an enhancement of heat transfer in some of the small clearances, and an impairment in others although overall, the enhancement is dominant leading to a better cooling. Secondly, the cross flow makes effective use of the small clearances between the key/keyway connections which increases the effective heat transfer area, hence increasing the cooling. Under the conditions of no cross flow, these areas remain largely inactive in heat transfer. The study shows that the cooling of the moderator is significantly enhanced by the cross flow perpendicular to the main cooling flow. (author)

  20. Development of a CVD silica coating for UK advanced gas-cooled nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Bennett, M.J.; Houlton, M.R.; Moore, D.A.; Foster, A.I.; Swidzinski, M.A.M.

    1983-04-01

    Vapour deposited silica coatings could extend the life of the 20% Cr/25% Ni niobium stabilised (20/25/Nb) stainless steel fuel cladding of the UK advanced gas cooled reactors. A CVD coating process developed originally to be undertaken at atmospheric pressure has now been adapted for operation at reduced pressure. Trials on the LP CVD process have been pursued to the production scale using commercial equipment. The effectiveness of the LP CVD silica coatings in providing protection to 20/25/Nb steel surfaces against oxidation and carbonaceous deposition has been evaluated. (author)

  1. Safety design features for current UK advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yellowlees, J. M.; Cobb, E. C. [Nuclear Power Co. (Risley) Ltd. (UK)

    1981-01-15

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed.

  2. Safety design features for current UK advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.; Cobb, E.C.

    1981-01-01

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed

  3. Microscopical examination of carbon deposits formed in the Windscale advanced gas cooled reactor

    International Nuclear Information System (INIS)

    Livesey, D.J.; Chatwin, W.H.; Pearce, J.H.

    1980-12-01

    Methods are described of sampling and examining carbon deposits on fuel cladding in the Windscale advanced gas-cooled reactor. Deposition is observed on fuel cladding in both the reactor core and experimental loops in carbon dioxide coolants containing various amounts of carbon monoxide and methane. Deposit distribution over the cladding surface indicated that nucleation is dependent on local surface conditions. Microscopical examination showed that deposit thickness increases by carbon filament growth into the coolant gas stream and that the process can be markedly influenced by metallic impurities. There is evidence that nickel can play a particularly significant role in deposition in loop experiments but similar effects have not been observed in the reactor core. (author)

  4. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  5. The rate of diffusion into advanced gas cooled reactor moderator bricks: an equivalent cylinder model

    International Nuclear Information System (INIS)

    Kyte, W.S.

    1980-01-01

    The graphite moderator bricks which make up the moderator of an advanced gas-cooled nuclear reactor (AGR) are of many different and complex shapes. Many physico-chemical processes that occur within these porous bricks include a diffusional step and thus to model these processes it is necessary to solve the diffusion equation (with chemical reaction) in a porous medium of complex shape. A finite element technique is applied to calculating the rate at which nitrogen diffuses into and out of the porous moderator graphite during operation of a shutdown procedure for an AGR. However, the finite element method suffers from several disadvantages that undermine its general usefulness for calculating rates of diffusion in AGR moderator cores. A model which overcomes some of these disadvantages is presented (the equivalent cylinder model) and it is shown that this gives good results for a variety of different boundary and initial conditions

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

    International Nuclear Information System (INIS)

    Mella, R.; Wenman, M.R.

    2013-01-01

    Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

  7. Multicycle Optimization of Advanced Gas-Cooled Reactor Loading Patterns Using Genetic Algorithms

    International Nuclear Information System (INIS)

    Ziver, A. Kemal; Carter, Jonathan N.; Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Overton, Richard S.

    2003-01-01

    A genetic algorithm (GA)-based optimizer (GAOPT) has been developed for in-core fuel management of advanced gas-cooled reactors (AGRs) at HINKLEY B and HARTLEPOOL, which employ on-load and off-load refueling, respectively. The optimizer has been linked to the reactor analysis code PANTHER for the automated evaluation of loading patterns in a two-dimensional geometry, which is collapsed from the three-dimensional reactor model. GAOPT uses a directed stochastic (Monte Carlo) algorithm to generate initial population members, within predetermined constraints, for use in GAs, which apply the standard genetic operators: selection by tournament, crossover, and mutation. The GAOPT is able to generate and optimize loading patterns for successive reactor cycles (multicycle) within acceptable CPU times even on single-processor systems. The algorithm allows radial shuffling of fuel assemblies in a multicycle refueling optimization, which is constructed to aid long-term core management planning decisions. This paper presents the application of the GA-based optimization to two AGR stations, which apply different in-core management operational rules. Results obtained from the testing of GAOPT are discussed

  8. Genetic algorithms and artificial neural networks for loading pattern optimisation of advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ziver, A.K. E-mail: a.k.ziver@imperial.ac.uk; Pain, C.C; Carter, J.N.; Oliveira, C.R.E. de; Goddard, A.J.H.; Overton, R.S

    2004-03-01

    A non-generational genetic algorithm (GA) has been developed for fuel management optimisation of Advanced Gas-Cooled Reactors, which are operated by British Energy and produce around 20% of the UK's electricity requirements. An evolutionary search is coded using the genetic operators; namely selection by tournament, two-point crossover, mutation and random assessment of population for multi-cycle loading pattern (LP) optimisation. A detailed description of the chromosomes in the genetic algorithm coded is presented. Artificial Neural Networks (ANNs) have been constructed and trained to accelerate the GA-based search during the optimisation process. The whole package, called GAOPT, is linked to the reactor analysis code PANTHER, which performs fresh fuel loading, burn-up and power shaping calculations for each reactor cycle by imposing station-specific safety and operational constraints. GAOPT has been verified by performing a number of tests, which are applied to the Hinkley Point B and Hartlepool reactors. The test results giving loading pattern (LP) scenarios obtained from single and multi-cycle optimisation calculations applied to realistic reactor states of the Hartlepool and Hinkley Point B reactors are discussed. The results have shown that the GA/ANN algorithms developed can help the fuel engineer to optimise loading patterns in an efficient and more profitable way than currently available for multi-cycle refuelling of AGRs. Research leading to parallel GAs applied to LP optimisation are outlined, which can be adapted to present day LWR fuel management problems.

  9. Experience of the remote dismantling of the Windscale advanced gas-cooled reactor and Windscale pile chimneys

    International Nuclear Information System (INIS)

    Wright, E.M.

    1993-01-01

    This paper gives brief descriptions of some of the remote dismantling work and equipment used on two large decommissioning projects: the BNFL Windscale Pile Chimneys Project (remote handling machine, waste packaging machine, remotely controlled excavator, remotely controlled demolition machine) and the AEA Windscale Advanced Gas-cooled Reactor Project (remote dismantling machine, operational waste, bulk removal techniques, semi-remote cutting operations)

  10. Structural instabilities of high temperature alloys and their use in advanced high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Schuster, H.; Ennis, P.J.; Nickel, H.; Czyrska-Filemonowicz, A.

    1989-01-01

    High-temperature, iron-nickel and nickel based alloys are the candidate heat exchanger materials for advanced high temperature gas-cooled reactors supplying process heat for coal gasification, where operation temperatures can reach 850-950 deg. C and service lives of more than 100,000 h are necessary. In the present paper, typical examples of structural changes which occur in two representative alloys (Alloy 800 H, Fe-32Ni-20Cr and Alloy 617, Ni-22Cr-12Co-9Mo-1Al) during high temperature exposure will be given and the effects on the creep rupture properties discussed. At service temperatures, precipitation of carbides occurs which has a significant effect on the creep behaviour, especially in the early stages of creep when the precipitate particles are very fine. During coarsening of the carbides, carbides at grain boundaries restrict grain boundary sliding which retards the development of creep damage. In the service environments, enhanced carbide precipitation may occur due to the ingress of carbon from the environment (carburization). Although the creep rate is not adversely affected, the ductility of the carburized material at low and intermediate temperatures is very low. During simulated service exposures, the formation of surface corrosion scales, the precipitation of carbides and the formation of internal oxides below the surface leads to depletion of the matrix in the alloying elements involved in the corrosion processes. In thin-walled tubes the depletion of Cr due to Cr 2 O 3 formation on the surface can lead to a loss of creep strength. An additional depletion effect resulting from environmental-metal reactions is the loss of carbon (decarburization) which may occur in specific environments. The compositions of the cooling gases which decarburize the material have been determined; they are to be avoided during reactor operation

  11. Remote handling equipment for the decommissioning of the Windscale Advanced Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Barker, A.; Birss, I.R.; Fish, G.

    1984-01-01

    A decision to decommission the Windscale Advanced Gas Cooled Reactor was taken shortly after reactor shutdown in 1981. The fuel has now been discharged and the decommissioning programme will last about 10-12 years. The paper describes the programme and objectives and deals with methods of handling and disposing of the radioactive waste material. The main new facility required is a Waste Packaging Building adjacent to the existing reactor in which the waste boxes will be filled, active waste encapsulated in concrete and the boxes cleaned, swabbed and monitored to comply with IAEA transport regulations. The handling machine concept and features are described. The assaying and packaging of the waste material, the control of box movement and the process of concrete encapsulation is described. The paper concludes with a description of the development programme to support the Project. The tasks include a study of cutting techniques, production and control of dust and smoke, viewing and lighting methods, filtration, decontamination and fixing of contamination

  12. Sensitivity analysis of an Advanced Gas-cooled Reactor control rod model

    International Nuclear Information System (INIS)

    Scott, M.; Green, P.L.; O’Driscoll, D.; Worden, K.; Sims, N.D.

    2016-01-01

    Highlights: • A model was made of the AGR control rod mechanism. • The aim was to better understand the performance when shutting down the reactor. • The model showed good agreement with test data. • Sensitivity analysis was carried out. • The results demonstrated the robustness of the system. - Abstract: A model has been made of the primary shutdown system of an Advanced Gas-cooled Reactor nuclear power station. The aim of this paper is to explore the use of sensitivity analysis techniques on this model. The two motivations for performing sensitivity analysis are to quantify how much individual uncertain parameters are responsible for the model output uncertainty, and to make predictions about what could happen if one or several parameters were to change. Global sensitivity analysis techniques were used based on Gaussian process emulation; the software package GEM-SA was used to calculate the main effects, the main effect index and the total sensitivity index for each parameter and these were compared to local sensitivity analysis results. The results suggest that the system performance is resistant to adverse changes in several parameters at once.

  13. Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

    Science.gov (United States)

    Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

    2018-05-01

    The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (steel cladding is retained despite He2+ implantation.

  14. Mitigating the Risk of Stress Corrosion of Austenitic Stainless Steels in Advanced Gas Cooled Reactor Boilers

    International Nuclear Information System (INIS)

    Bull, A.; Owen, J.; Quirk, G.; G, Lewis; Rudge, A.; Woolsey, I.S.

    2012-09-01

    Advanced Gas-Cooled Reactors (AGRs) operated in the UK by EDF Energy have once-through boilers, which deliver superheated steam at high temperature (∼500 deg. C) and pressure (∼150 bar) to the HP turbine. The boilers have either a serpentine or helical geometry for the tubing of the main heat transfer sections of the boiler and each individual tube is fabricated from mild steel, 9%Cr1%Mo and Type 316 austenitic stainless steel tubing. Type 316 austenitic stainless steel is used for the secondary (final) superheater and steam tailpipe sections of the boiler, which, during normal operation, should operate under dry, superheated steam conditions. This is achieved by maintaining a specified margin of superheat at the upper transition joint (UTJ) between the 9%Cr1%Mo primary superheater and the Type 316 secondary superheater sections of the boiler. Operating in this mode should eliminate the possibility of stress corrosion cracking of the Type 316 tube material on-load. In recent years, however, AGRs have suffered a variety of operational problems with their boilers that have made it difficult to maintain the specified superheat margin at the UTJ. In the case of helical boilers, the combined effects of carbon deposition on the gas side and oxide deposition on the waterside of the tubing have resulted in an increasing number of austenitic tubes operating with less than the specified superheat margin at the UTJ and hence the possibility of wetting the austenitic section of the boiler. Some units with serpentine boilers have suffered creep-fatigue damage of the high temperature sections of the boiler, which currently necessitates capping the steam outlet temperature to prevent further damage. The reduction in steam outlet temperature has meant that there is an increased risk of operation with less than the specified superheat margin at the UTJ and hence stress corrosion cracking of the austenitic sections of the boiler. In order to establish the risk of stress

  15. Impingement jet cooling in gas turbines

    CERN Document Server

    Amano, R S

    2014-01-01

    Due to the requirement for enhanced cooling technologies on modern gas turbine engines, advanced research and development has had to take place in field of thermal engineering. Impingement jet cooling is one of the most effective in terms of cooling, manufacturability and cost. This is the first to book to focus on impingement cooling alone.

  16. Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1988-01-01

    Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized

  17. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  18. Nondestructive testing of welds in steam generators for advanced gas cooled reactors at Heyshamm II and Torness

    International Nuclear Information System (INIS)

    Parkin, K.; Bainbridge, A.; Carver, K.; Hammell, R.; Lack, B.J.

    1985-01-01

    The paper concerns non-destructive testing (NDT) of welds in advanced gas cooled steam generators for Heysham II and Torness nuclear power stations. A description is given of the steam generator. The selection of NDT techniques is also outlined, including the factors considered to ascertain the viability of a technique. Examples are given of applied NDT methods which match particular fabrication processes; these include: microfocus radiography, ultrasonic testing of austenitic tube butt welds, gamma-ray isotope projection system, surface crack detection, and automated radiography. Finally, future trends in this field of NDT are highlighted. (UK)

  19. Gas cooled reactors

    International Nuclear Information System (INIS)

    Kojima, Masayuki.

    1985-01-01

    Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

  20. Progress in the development of tooling and dismantling methodologies for the Windscale advanced gas cooled reactor (WAGR)

    International Nuclear Information System (INIS)

    Cross, M.T.; Wareing, M.I.; Dixon, C.

    1998-01-01

    Decommissioning of the Windscale Advanced Gas-Cooled Reactor (WAGR) is a major UK reactor decommissioning project co-funded by the UK Government, the European Commission and Magnox Electric. WAGR was a CO 2 cooled, graphite moderated reactor which served as a test bed for the development of Advanced Gas-Cooled Reactor technology in the UK. It operated from 1963 until shutdown in 1981. AEA Technology plc are currently the Managing Agents on behalf of UKAEA for the WAGR decommissioning project and are responsible for the co-ordination of the project up to the point when the contents of the reactor core and associated radioactive materials are removed and either disposed of or packaged for disposal at some time in the future. Decommissioning has progressed to the point where the reactor has been dismantled down to the level of the hot gas collection manifold with the removal of the top biological shield, the refuelling standpipes and the top section of the reactor pressure vessel. The 4 heat exchangers have also been removed and committed to shallow land burial. This paper describes the work carried out by AEA Technology under separate contracts of UKAEA in developing some of the equipment and deployment methods for the next phase of active operations required in preparation for the dismantling of the core structure. Most recent work has concentrated on the development of specialist tooling for removal of items of operational waste stored within the reactor core, equipment for cutting and removal of the highly radioactive stainless steel 'loop' pressure tubes, diamond wire cutting equipment for sectioning large diameter pipework, and equipment for dismantling the reactor neutron shield. The paper emphasises the process of adaptation and extension of existing technologies for cost-effective application in the decommissioning environment, the need for adequate forward planning of decommissioning methodologies together with large-scale 'mock-up' testing of equipment to

  1. Estimating the occurrence of foreign material in Advanced Gas-cooled Reactors: A Bayesian Monte Carlo approach

    International Nuclear Information System (INIS)

    Mason, Paolo

    2014-01-01

    Highlights: • The amount of a specific type of foreign material found in UK AGRs has been estimated. • The estimate is based on very few instances of detection in numerous inspections. • A Bayesian Monte Carlo approach was used. • The study supports safety case claims on coolant flow impairment. • The methodology is applicable to any inspection campaign on any plant system. - Abstract: The current occurrence of a particular sort of foreign material in eight UK Advanced Gas-cooled Reactors has been estimated by means of a parametric approach. The study includes both variability, treated in analytic fashion via the combination of standard probability distributions, and the uncertainty in the parameters of the model of choice, whose posterior distribution was inferred in Bayesian fashion by means of a Monte Carlo route consisting in the conditional acceptance of sets of model parameters drawn from a prior distribution based on engineering judgement. The model underlying the present study specifically refers to the re-loading and inspection routines of UK Advanced Gas-cooled Reactors. The approach to inference here presented, however, is of general validity and can be applied to the outcome of any inspection campaign on any plant system, and indeed to any situation in which the outcome of a stochastic process is more easily simulated than described by a probability density or mass function

  2. Multi-purpose nuclear heat source for advanced gas-cooled reactor plants

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1993-01-01

    Nuclear power has the potential to be the ultimate green technology in that it could eliminate the need for burning fossil fuels with their polluting combustion products and greenhouse gases. This view is shared by many technologists, but it may be a generation before the public becomes convinced, and that will involve overcoming many safety, institutional, financial, and technical impediments. This paper addresses only the latter topic; a major theme being that for nuclear power to truly be a green technology and significantly benefit society, it must meet the needs of the full energy spectrum. Specifically, it must satisfy energy needs beyond just the electricity generating sector by today's nuclear plants. By virtue of its high temperature capability, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is the only type of reactor that has the potential to meet the wide range of energy needs that will emerge in the future. This paper discusses the nuclear heat source that gives the MHTGR multi-purpose capability, which is recognized today, but will not be implemented until early in the next century

  3. Development of advanced fabrication technology for high-temperature gas-cooled reactor fuel. Reduction of coating failure fraction

    International Nuclear Information System (INIS)

    Minato, Kazuo; Kikuchi, Hironobu; Fukuda, Kousaku; Tobita, Tsutomu; Yoshimuta, Sigeharu; Suzuki, Nobuyuki; Tomimoto, Hiroshi; Nishimura, Kazuhisa; Oda, Takafumi

    1998-11-01

    The advanced fabrication technology for high-temperature gas-cooled reactor fuel has been developed to reduce the coating failure fraction of the fuel particles, which leads to an improvement of the reactor safety. The present report reviews the results of the relevant work. The mechanisms of the coating failure of the fuel particles during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the coating failure fraction of the fuel. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the fuel was improved outstandingly. (author)

  4. Preparation for Future Defuelling and Decommissioning Works on EDF Energy's UK Fleet of Advanced Gas Cooled Reactors

    International Nuclear Information System (INIS)

    Bryers, John; Ashmead, Simon

    2016-01-01

    EDF Energy/Nuclear Generation is the owner and operator of 14 Advanced Gas cooled Reactors (AGR) and one Pressurised Water Reactor (PWR), on 8 nuclear stations in the UK. EDF Energy/Nuclear Generation is responsible for all the activities associated with the end of life of its nuclear installations: de-fuelling, decommissioning and waste management. As the first AGR is forecast to cease generation within 10 years, EDF Energy has started planning for the decommissioning. This paper covers: - broad outline of the technical strategy and arrangements for future de-fuelling and decommissioning works on the UK AGR fleet, - high level strategic drivers and alignment with wider UK nuclear policy, - overall programme of preparation and initial works, - technical approaches to be adopted during decommissioning. (authors)

  5. Advanced Gas Cooled Reactor Materials Program. Reducing helium impurity depletion in HTGR materials testing

    International Nuclear Information System (INIS)

    Baldwin, D.H.

    1984-08-01

    Moisture depletion in HTGR materials testing rigs has been empirically studied in the GE High Temperature Reactor Materials Testing Laboratory (HTRMTL). Tests have shown that increased helium flow rates and reduction in reactive (oxidizable) surface area are effective means of reducing depletion. Further, a portion of the depletion has been shown to be due to the presence of free C released by the dissociation of CH 4 . This depletion component can be reduced by reducing the helium residence time (increasing the helium flow rate) or by reducing the CH 4 concentration in the test gas. Equipment modifications to reduce depletion have been developed, tested, and in most cases implemented in the HTRMTL to date. These include increasing the Helium Loop No. 1 pumping capacity, conversion of metallic retorts and radiation shields to alumina, isolation of thermocouple probes from the test gas by alumina thermowells, and substitution of non-reactive Mo-TZM for reactive metallic structural components

  6. Behaviour of particles in a commercial advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    Garland, J.A.; Wells, A.C.; Hedgecock, J.B.

    1985-01-01

    Certain hypothetical fault conditions cause fission products to escape from a fraction of the fuel to an intact coolant circuit. A significant fraction of the activity, including iodine isotopes, is expected to attach to small particles suspended in the gas coolant, and the fate of the particles may influence the fraction of the activity available to escape to the environment with the small amount of gas that leaks continuously from the coolant circuit. A series of experiments has provided an understanding of the behaviour of such particles. Tracer particles of 0.6, 2, 5 and 17 μm diameter, labelled with 59 Fe, were dispersed as aerosols in the reactor coolant, and the subsequent variation of concentration was observed by measurement of a sequence of filter samples of the coolant gas. The changes in concentration were influenced by mixing processes, but showed clearly that loss processes reduced the burden in the coolant by two or three orders of magnitude within 3 h. The concentration did not follow a simple exponential decrease. Small particles deposited more rapidly than the largest size studied. These observations imply that particles both impact onto, and also bounce and resuspend from, the internal surfaces of the coolant circuit. Although the physical mechanisms of the particle-surface interaction cannot be described in detail, the results clearly demonstrate a large benefit due to deposition reducing the amount of circulating activity. The quantity of particle-borne activity available for escape with leaking coolant during 24 h following a release from fuel is reduced by a factor ranging from several hundred to a few thousand. (author)

  7. Boiler referruling on the Hartlepool and Heysham 1 advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    Newell, J.E.

    1988-01-01

    The Hartlepool and Heysham I reactors each use eight cylindrical boilers having nineteen rows of helical tubes. The advantages of this design are partially offset by the relatively poor radial gas mixing. Some rows of tubing may have an imbalance between heat input from the gas and the flow of feedwater. causing a temperature profile at the upper transition joints. The thermal/hydraulic behaviour meant that the metallurgical constraints limited output. Analysis of the behaviour of these boilers required a new two-dimensional mathematical model, known as PODMIX. This describes the thermal hydraulics in each of the rows of tubing and also in the gas between the rows. Not all of the parameters for the model can be determined from first principles. However, two out of the thirty two pods have thermocouples at some of the upper transition joints and these made back calculation possible. In order to translate this model to other boiler pods, a novel thermocouple rake system was designed for sampling superheated steam temperatures in selected tubes. A result of this analysis was to show that different, individual ferrule patterns were needed for each pod. The characteristics could, in general, best be met using twin orifice ferrules. Unfortunately, the installed system did not permit the replacement of orifices, so that a completely new system had to be developed. In the course of designing this, the opportunity was taken to over come susceptibilities to erosion/corrosion and crevice corrosion. Removal of the old ferrules and replacement with the new ones necessitated the development of high precision, programmable machines to operate under difficult site conditions. These carried out drilling, boring, grinding and polishing operations as well as making face welds and tube bore welds. Modifications have already achieved substantial improvements in performance and output, but an extended, iterative programme still lies ahead. (author)

  8. Evaluation of materials' corrosion and chemistry issues for advanced gas cooled reactor steam generators using full scale plant simulations

    International Nuclear Information System (INIS)

    Woolsey, I.S.; Rudge, A.J.; Vincent, D.J.

    1998-01-01

    Advanced Gas Cooled Reactors (AGRS) employ once-through steam Generators of unique design to provide steam at approximately 530 degrees C and 155 bar to steam turbines of similar design to those of fossil plants. The steam generators are highly compact, and have either a serpentine or helical tube geometry. The tubes are heated on the outside by hot C0 2 gas, and steam is generated on the inside of the tubes. Each individual steam generator tube consists of a carbon steel feed and primary economiser section, a 9%Cr steel secondary economiser, evaporator and primary superheater, and a Type 316L austenitic stainless steel secondary superheater, all within a single tube pass. The multi-material nature of the individual tube passes, the need to maintain specific thermohydraulic conditions within the different material sections, and the difficulties of steam generator inspection and repair, have required extensive corrosion-chemistry test programmes to ensure waterside corrosion does not present a challenge to their integrity. A major part of these programmes has been the use of a full scale steam generator test facility capable of simulating all aspects of the waterside conditions which exist in the plant. This facility has been used to address a wide variety of possible plant drainage/degradation processes. These include; single- and two-phase flow accelerated corrosion of carbon steel, superheat margins requirements and the stress-corrosion behaviour of the austenitic superheaters, on-load corrosion of the evaporator materials, and iron transport and oxide deposition behaviour. The paper outlines a number of these, and indicates how they have been of value in helping to maintain reliable operation of the plant. (author)

  9. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    1974-01-01

    The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines [fr

  10. An advanced three-dimensional simulation system for safety analysis of gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lapins, Janis

    2016-07-01

    The neutron transport programme TORT-TD that solves the neutron transport equation in discrete ordinates for stationary as well as transient problems is used for neutronics calculations. The transient solution of the neutron transport equation is performed by making use of a time-dependent neutron source, xenon/iodine dynamics are implemented as well. The programme ATTICA{sup 3D} applies the porous medium approach for flow in packed beds according to Ergun. This approach uses a quasi-steady state formulation for the momentum equation while time dependent formulations are employed for mass conservation, and energy conservation for both, the solid and gaseous phase. For spatial discretisation of the conservation equations, the finite volume method is used. For material properties, gas densities, heat transfer etc. a set of constitutive equations completes the set of differential equations. Time integration in ATTICA{sup 3D} is realised applying a modified Newton-Raphson method which linearizes and subsequently solves the set of equations. It can automatically adapt the time step width within user specified limits. Within this work, the mass and energy conservation equations are modified so that chemical reactions as consequence of water or air ingress can be simulated, i.e. mass sources for CO, CO{sub 2}, H{sub 2} and sinks for H{sub 2}O and O{sub 2} or heat sources and enthalpy transport. The heat generated by chemical reaction is either added to the solid or the gaseous phase. The corrosion rates were implemented according to experimental findings for fuel and reflector graphite. Steam or air might enter the primary circuit through a break in the steam generator or an opening of the primary circuit. Steam entering the core region will moderate neutrons, reduce the leakage and thereby increase power. The corrosion was validated for the NACOK experiment performed within the RAPHAEL project where temperature evolution under corrosion and total burn-off had to be

  11. Gas cooled leads

    International Nuclear Information System (INIS)

    Shutt, R.P.; Rehak, M.L.; Hornik, K.E.

    1993-01-01

    The intent of this paper is to cover as completely as possible and in sufficient detail the topics relevant to lead design. The first part identifies the problems associated with lead design, states the mathematical formulation, and shows the results of numerical and analytical solutions. The second part presents the results of a parametric study whose object is to determine the best choice for cooling method, material, and geometry. These findings axe applied in a third part to the design of high-current leads whose end temperatures are determined from the surrounding equipment. It is found that cooling method or improved heat transfer are not critical once good heat exchange is established. The range 5 5 but extends over a large of values. Mass flow needed to prevent thermal runaway varies linearly with current above a given threshold. Below that value, the mass flow is constant with current. Transient analysis shows no evidence of hysteresis. If cooling is interrupted, the mass flow needed to restore the lead to its initially cooled state grows exponentially with the time that the lead was left without cooling

  12. Gas hydrate cool storage system

    Science.gov (United States)

    Ternes, M.P.; Kedl, R.J.

    1984-09-12

    The invention presented relates to the development of a process utilizing a gas hydrate as a cool storage medium for alleviating electric load demands during peak usage periods. Several objectives of the invention are mentioned concerning the formation of the gas hydrate as storage material in a thermal energy storage system within a heat pump cycle system. The gas hydrate was formed using a refrigerant in water and an example with R-12 refrigerant is included. (BCS)

  13. Use of oxygen dosing to prevent flow accelerated corrosion in British Energy's Advanced Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Quirk, G.P.; Woolsey, I.S.; Rudge, A.

    2010-01-01

    Flow accelerated corrosion (FAC) was recognized as major threat to the carbon steel feed and economizer tubing of the once-through boilers of the UK's Advanced Gas-cooled Reactors (AGRs) following the observation of FAC damage of the boiler inlet orifice assemblies at two plants in 1977, and subsequent review of the likelihood of further damage elsewhere within the boilers of all AGRs. In most cases, replacement of susceptible tubing was not feasible; due to the inaccessibility of the boiler components within the reactor concrete pressure vessel. Preventing further FAC damage within the boilers therefore had to rely largely on changes to the boiler feedwater chemistry. Following extensive research programs carried out in the late 1970s and early 1980s two main feedwater chemistry regimes were adopted to suppress FAC in different AGRs. The four units found to be at greatest risk of FAC damage adopted an oxygen dosed All Volatile Treatment (AVT) regime during commissioning, while four other units retained the original deoxygenated ammonia dosed AVT regime, but with an increased feedwater pH. The deoxygenated ammonia dosed chemistry regime was also adopted in four AGR units subsequently built, which used 1%Cr0.5%Mo feed and economizer tubing in their once-through boilers. The oxygen dosed AVT chemistry regime adopted in four units having helical once-through boilers has proved highly effective in preventing FAC, with no evidence of damage after around 150,000 hours of operation. However, FAC damage was eventually found in some of the other units operating with a deoxygenated feedwater chemistry regime, in spite of having adopted an elevated feedwater pH. These units have now successfully converted to an oxygen dosed AVT feedwater chemistry regime to prevent further FAC damage, with the result that all 14 AGR reactors now operate with variants of the original oxygen dosed feedwater chemistry regime developed during the 1980s. The paper outlines the development of

  14. Gas cooled HTR

    International Nuclear Information System (INIS)

    Schweiger, F.

    1985-01-01

    In the He-cooled, graphite-moderated HTR with spherical fuel elements, the steam generator is fixed outside the pressure vessel. The heat exchangers are above the reactor level. The hot gases stream from the reactor bottom over the heat exchanger, through an annular space around the heat exchanger and through feed lines in the side reflector of the reactor back to its top part. This way, in case of shutdown there is a supplementary natural draught that helps the inner natural circulation (chimney draught effect). (orig./PW)

  15. Gas cooled traction drive inverter

    Science.gov (United States)

    Chinthavali, Madhu Sudhan

    2013-10-08

    The present invention provides a modular circuit card configuration for distributing heat among a plurality of circuit cards. Each circuit card includes a housing adapted to dissipate heat in response to gas flow over the housing. In one aspect, a gas-cooled inverter includes a plurality of inverter circuit cards, and a plurality of circuit card housings, each of which encloses one of the plurality of inverter cards.

  16. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  17. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  18. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    Maturana, Roberto Hernan

    2008-01-01

    At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design. [es

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

    International Nuclear Information System (INIS)

    Garcia, Rosales; Fajardo, Garcia; Curbelo, Perez; Oliva, Munoz; Hernandez, Garcia; Castells, Escriva; Abanades

    2011-01-01

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

  20. Core catcher cooling for a gas-cooled fast breeder

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Schretzmann, K.

    1976-01-01

    Water, molten salts, and liquid metals are under discussion as coolants for the core catcher of a gas-cooled fast breeder. The authors state that there is still no technically mature method of cooling a core melt. However, the investigations carried out so far suggest that there is a solution to this problem. (RW/AK) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  2. Design codes for gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-12-01

    High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  3. Gas fired advanced turbine system

    Science.gov (United States)

    Lecren, R. T.; White, D. J.

    The basic concept thus derived from the Ericsson cycle is an intercooled, recuperated, and reheated gas turbine. Theoretical performance analyses, however, showed that reheat at high turbine rotor inlet temperatures (TRIT) did not provide significant efficiency gains and that the 50 percent efficiency goal could be met without reheat. Based upon these findings, the engine concept adopted as a starting point for the gas-fired advanced turbine system is an intercooled, recuperated (ICR) gas turbine. It was found that, at inlet temperatures greater than 2450 F, the thermal efficiency could be maintained above 50%, provided that the turbine cooling flows could be reduced to 7% of the main air flow or lower. This dual and conflicting requirement of increased temperatures and reduced cooling will probably force the abandonment of traditional air cooled turbine parts. Thus, the use of either ceramic materials or non-air cooling fluids has to be considered for the turbine nozzle guide vanes and turbine blades. The use of ceramic components for the proposed engine system is generally preferred because of the potential growth to higher temperatures that is available with such materials.

  4. Proceedings: Cooling tower and advanced cooling systems conference

    International Nuclear Information System (INIS)

    1995-02-01

    This Cooling Tower and Advanced Cooling Systems Conference was held August 30 through September 1, 1994, in St. Petersburg, Florida. The conference was sponsored by the Electric Power Research Institute (EPRI) and hosted by Florida Power Corporation to bring together utility representatives, manufacturers, researchers, and consultants. Nineteen technical papers were presented in four sessions. These sessions were devoted to the following topics: cooling tower upgrades and retrofits, cooling tower performance, cooling tower fouling, and dry and hybrid systems. On the final day, panel discussions addressed current issues in cooling tower operation and maintenance as well as research and technology needs for power plant cooling. More than 100 people attended the conference. This report contains the technical papers presented at the conference. Of the 19 papers, five concern cooling tower upgrades and retrofits, five to cooling tower performance, four discuss cooling tower fouling, and five describe dry and hybrid cooling systems. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  5. Natural gas cooling: Part of the solution

    International Nuclear Information System (INIS)

    Jones, D.R.

    1992-01-01

    This paper reviews and compares the efficiencies and performance of a number of gas cooling systems with a comparable electric cooling system. The results show that gas cooling systems compare favorably with the electric equivalents, offering a new dimension to air conditioning and refrigeration systems. The paper goes on to compare the air quality benefits of natural gas to coal or oil-burning fuel systems which are used to generate the electricity for the electric cooling systems. Finally, the paper discusses the regulatory bias that the author feels exists towards the use of natural gas and the need for modification in the existing regulations to provide a 'level-playing field' for the gas cooling industry

  6. Heating up the gas cooling market

    International Nuclear Information System (INIS)

    Watt, G.

    2001-01-01

    Gas cooling is an exciting technology with a potentially bright future. It comprises the production of cooling (and heating) in buildings and industry, by substituting environmentally-friendlier natural gas or LPG over predominantly coal-fired electricity in air conditioning equipment. There are currently four established technologies using gas to provide cooling energy or conditioned air. These are: absorption, both direct gas-fired and utilising hot water or steam; gas engine driven vapour compression (GED); cogeneration, with absorption cooling driven by recovered heat; and desiccant systems. The emergence of gas cooling technologies has been, and remains, one of evolution rather than revolution. However, further development of the technology has had a revolutionary effect on the performance, reliability and consumer acceptability of gas cooling products. Developments from world-renowned manufacturers such as York, Hitachi, Robur and Thermax have produced a range of absorption equipment variously offering: the use of 100 percent environmentally-friendly refrigerants, with zero global warming potential; the ideal utilisation of waste heat from cogeneration systems; a reduction in electrical distribution and stand-by generation capacity; long product life expectancy; far less noise and vibration; performance efficiency maintained down to about 20 percent of load capacity; and highly automated and low-cost maintenance. It is expected that hybrid systems, that is a mixture of gas and electric cooling technologies, will dominate the future market, reflecting the uncertainty in the electricity market and the prospects of stable future gas prices

  7. Emergency cooling system for a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Cook, R.K.; Burylo, P.S.

    1975-01-01

    The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

  8. Hydrogen production by high-temperature gas-cooled reactor. Conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki; Hara, Teruo; Kato, Ryoma; Kunitomi, Kazuhiko

    2008-01-01

    Nuclear hydrogen production is necessary in an anticipated hydrogen society that demands a massive quantity of hydrogen without economic disadvantage. Japan Atomic Energy Agency (JAEA) has launched the conceptual design study of a hydrogen production system with a near-term plan to connect it to Japan's first high-temperature gas-cooled reactor HTTR. The candidate hydrogen production system is based on the thermochemical water-splitting iodine sulphur (IS) process.The heat of 10 MWth at approximately 900degC, which can be provided by the secondary helium from the intermediate heat exchanger of the HTTR, is the energy input to the hydrogen production system. In this paper, we describe the recent progresses made in the conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system. A new concept of sulphuric acid decomposer is proposed. This involves the integration of three separate functions of sulphuric acid decomposer, sulphur trioxide decomposer, and process heat exchanger. A new mixer-settler type of Bunsen reactor is also designed. This integrates three separate functions of Bunsen reactor, phase separator, and pump. The new concepts are expected to result in improved economics through construction and operation cost reductions because the number of process equipment and complicated connections between the equipment has been substantially reduced. (author)

  9. Characterizing the effects of elevated temperature on the air void pore structure of advanced gas-cooled reactor pressure vessel concrete using x-ray computed tomography

    Directory of Open Access Journals (Sweden)

    Withers P.J.

    2013-07-01

    Full Text Available X-ray computed tomography (X-ray CT has been applied to nondestructively characterise changes in the microstructure of a concrete used in the pressure vessel structure of Advanced Gas-cooled Reactors (AGR in the UK. Concrete specimens were conditioned at temperatures of 105 °C and 250 °C, to simulate the maximum thermal load expected to occur during a loss of coolant accident (LOCA. Following thermal treatment, these specimens along with an unconditioned control sample were characterised using micro-focus X-ray CT with a spatial resolution of 14.6 microns. The results indicate that the air void pore structure of the specimens experienced significant volume changes as a result of the increasing temperature. The increase in the porous volume was more prevalent at 250 °C. Alterations in air void size distributions were characterized with respect to the unconditioned control specimen. These findings appear to correlate with changes in the uni-axial compressive strength of the conditioned concrete.

  10. Compatibility of gas turbine materials with steam cooling

    Energy Technology Data Exchange (ETDEWEB)

    Desai, V.; Tamboli, D.; Patel, Y. [Univ. of Central Florida, Orlando, FL (United States)

    1995-10-01

    Gas turbines had been traditionally used for peak load plants and remote locations as they offer advantage of low installation costs and quick start up time. Their use as a base load generator had not been feasible owing to their poor efficiency. However, with the advent of gas turbines based combined cycle plants (CCPs), continued advances in efficiency are being made. Coupled with ultra low NO{sub x} emissions, coal compatibility and higher unit output, gas turbines are now competing with conventional power plants for base load power generation. Currently, the turbines are designed with TIT of 2300{degrees}F and metal temperatures are maintained around 1700{degrees}F by using air cooling. New higher efficiency ATS turbines will have TIT as high as 2700{degrees}F. To withstand this high temperature improved materials, coatings, and advances in cooling system and design are warranted. Development of advanced materials with better capabilities specifically for land base applications are time consuming and may not be available by ATS time frame or may prove costly for the first generation ATS gas turbines. Therefore improvement in the cooling system of hot components, which can take place in a relatively shorter time frame, is important. One way to improve cooling efficiency is to use better cooling agent. Steam as an alternate cooling agent offers attractive advantages because of its higher specific heat (almost twice that of air) and lower viscosity.

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

    International Nuclear Information System (INIS)

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Collins, J.L.

    2004-01-01

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

  13. Improvement of Cooling Technology through Atmosphere Gas Management

    Energy Technology Data Exchange (ETDEWEB)

    Renard, Michel; Dosogne, Edgaar; Crutzen, Jean Pierre; Raick, Jean Mare [DREVER INTERNATIONAL S.A., Liege (Belgium); Ji, Ma Jia; Jun, Lv; Zhi, Ma Bing [SHOUGANG Cold Rolling Mill Headquarter, Beijin (China)

    2009-12-15

    The production of advanced high strength steels requires the improvement of cooling technology. The use of high cooling rates allows relatively low levels of expensive alloying additions to ensure sufficient hardenability. In classical annealing and hot-dip galvanizing lines a mixing station is used to provide atmosphere gas containing 3-5% hydrogen and 97-95% nitrogen in the various sections of the furnace, including the rapid cooling section. Heat exchange enhancement in this cooling section can be insured by the increased hydrogen concentration. Driver international developed a patented improvement of cooling technology based on the following features: pure hydrogen gas is injected only in the rapid cooling section whereas the different sections of the furnace are supplied with pure nitrogen gas: the control of flows through atmosphere gas management allows to get high hydrogen concentration in cooling section and low hydrogen content in the other furnace zones. This cooling technology development insures higher cooling rates without additional expensive hydrogen gas consumption and without the use of complex sealing equipment between zones. In addition reduction in electrical energy consumption is obtained. This atmosphere control development can be combined with geometrical design improvements in order to get optimised cooling technology providing high cooling rates as well as reduced strip vibration amplitudes. Extensive validation of theoretical research has been conducted on industrial lines. New lines as well as existing lines, with limited modifications, can be equipped with this new development. Up to now this technology has successfully been implemented on 6 existing and 7 new lines in Europe and Asia.

  14. Processing of coke oven gas. Primary gas cooling

    Energy Technology Data Exchange (ETDEWEB)

    Ullrich, H [Otto (C.) und Co. G.m.b.H., Bochum (Germany, F.R.)

    1976-11-01

    The primary cooler is an indispensable part of all byproduct processing plants. Its purpose is to cool the raw gas from the coke oven battery and to remove the accompanying water vapor. The greater part of the cooling capacity is utilized for the condensation of water vapor and only a small capacity is needed for the gas cooling. Impurities in the gas, like naphthalene, tar and solid particles, necessitate a special design in view of the inclination to dirt accumulation. Standard types of direct and indirect primary gas coolers are described, with a discussion of their advantages and disadvantages.

  15. The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474

    Energy Technology Data Exchange (ETDEWEB)

    Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

    2012-07-01

    The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the

  16. Dimethylamine as a Replacement for Ammonia Dosing in the Secondary Circuit of an Advanced Gas-Cooled Reactor (AGR) Power Station

    International Nuclear Information System (INIS)

    Armstrong, C.; Mitchell, M.; Bull, A.; Quirk, G.P.; Rudge, A.

    2012-09-01

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

  17. Advanced technology for aero gas turbine components

    Energy Technology Data Exchange (ETDEWEB)

    1987-09-01

    The Symposium is aimed at highlighting the development of advanced components for new aero gas turbine propulsion systems in order to provide engineers and scientists with a forum to discuss recent progress in these technologies and to identify requirements for future research. Axial flow compressors, the operation of gas turbine engines in dust laden atmospheres, turbine engine design, blade cooling, unsteady gas flow through the stator and rotor of a turbomachine, gear systems for advanced turboprops, transonic blade design and the development of a plenum chamber burner system for an advanced VTOL engine are among the topics discussed.

  18. Hot gas path component cooling system

    Science.gov (United States)

    Lacy, Benjamin Paul; Bunker, Ronald Scott; Itzel, Gary Michael

    2014-02-18

    A cooling system for a hot gas path component is disclosed. The cooling system may include a component layer and a cover layer. The component layer may include a first inner surface and a second outer surface. The second outer surface may define a plurality of channels. The component layer may further define a plurality of passages extending generally between the first inner surface and the second outer surface. Each of the plurality of channels may be fluidly connected to at least one of the plurality of passages. The cover layer may be situated adjacent the second outer surface of the component layer. The plurality of passages may be configured to flow a cooling medium to the plurality of channels and provide impingement cooling to the cover layer. The plurality of channels may be configured to flow cooling medium therethrough, cooling the cover layer.

  19. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

  20. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    1992-09-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

  1. Ship exhaust gas plume cooling

    NARCIS (Netherlands)

    Schleijpen, H.M.A.; Neele, P.P.

    2004-01-01

    The exhaust gas plume is an important and sometimes dominating contributor to the infrared signature of ships. Suppression of the infrared ship signatures has been studied by TNO for the Royal Netherlands Navy over considerable time. This study deals with the suppression effects, which can be

  2. French activities on gas cooled reactors

    International Nuclear Information System (INIS)

    Bastien, D.

    1996-01-01

    The gas cooled reactor programme in France originally consisted of eight Natural Uranium Graphite Gas Cooled Reactors (UNGG). These eight units, which are now permanently shutdown, represented a combined net electrical power of 2,375 MW and a total operational history of 163 years. Studies related to these reactors concern monitoring and dismantling of decommissioned facilities, including the development of methods for dismantling. France has been monitoring the development of HTRs throughout the world since 1979, when it halted its own HTR R and D programme. France actively participates in three CRPs set up by the IAEA. (author). 1 tab

  3. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  4. Thermohydraulic relationships for advanced water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  5. Estimation of gas turbine blades cooling efficiency

    NARCIS (Netherlands)

    Moskalenko, A.B.; Kozhevnikov, A.

    2016-01-01

    This paper outlines the results of the evaluation of the most thermally stressed gas turbine elements, first stage power turbine blades, cooling efficiency. The calculations were implemented using a numerical simulation based on the Finite Element Method. The volume average temperature of the blade

  6. Gas-cooled reactor technology: a bibliography

    International Nuclear Information System (INIS)

    Raleigh, H.D.

    1981-09-01

    Included are 3358 citations on gas-cooled reactor technology contained in the DOE Energy Data Base for the period January 1978 through June 1981. The citations include reports, journal articles, books, conference papers, patents, and monographs. Corporate, Personal Author, Subject, Contract Number, and Report Number Indexes are provided

  7. High temperature gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hosegood, S.B.; Lockett, G.E.

    1975-01-01

    For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

  8. Cooling process in separation devices of krypton gas

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, S; Sugimoto, K

    1975-06-11

    To prevent entry of impurities into purified gases and to detect leaks of heat exchanger in a separation and recovering device of krypton gas by means of liquefaction and distillation, an intermediate refrigerant having the same or slightly higher boiling point than that of gas to be cooled is used between the gas to be cooled (process gas) and refrigerant (nitrogen), and the pressure of the gas to be cooled is controlled to have a pressure higher than the intermediate refrigerant to cool the gas to be cooled.

  9. The modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Lutz, D.E.; Lipps, A.J.

    1984-01-01

    Due to relatively high operating temperatures, the gas-cooled reactor has the potential to serve a wide variety of energy applications. This paper discusses the energy applications which can be served by the modular HTGR, the magnitude of the potential markets, and the HTGR product cost incentives relative to fossil fuel competition. Advantages of the HTGR modular systems are presented along with a description of the design features and performance characteristics of the current reference HTGR modular systems

  10. Gas-cooled breeder reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

    1981-01-15

    The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

  11. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  12. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  13. CEA programme on gas cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Chapelot, Ph.; Gauthier, J.C.

    2002-01-01

    Future nuclear energy systems studies conducted by the CEA aim at investigating and developing promising technologies for future reactors, fuels and fuel cycles, for nuclear power to play a major part in sustainable energy policies. Reactors and fuel cycles are considered as integral parts of a nuclear system to be optimised as a whole. Major goals assigned to future nuclear energy systems are the following: reinforced economic competitiveness with other electricity generation means, with a special emphasis on reducing the investment cost; enhanced reliability and safety, through an improved management of reactor operation in normal and abnormal plant conditions; minimum production of long lived radioactive waste; resource saving through an effective and flexible use of the available resources of fissile and fertile materials; enhanced resistance to proliferation risks. The three latter goals are essential for the sustainability of nuclear energy in the long term. Additional considerations such as the potentialities for other applications than electricity generation (co-generation, production of hydrogen, sea water desalination) take on an increasing importance. Sustainability goals call for fast neutron spectra (to transmute nuclear waste and to breed fertile fuel) and for recycling actinides from the spent fuel (plutonium and minor actinides). New applications and economic competitiveness call for high temperature technologies (850 deg C), that afford high conversion efficiencies and hence less radioactive waste production and discharged heat. These orientations call for breakthroughs beyond light water reactors. Therefore, as a result of a screening review of candidate technologies, the CEA has selected an innovative concept of high temperature gas cooled reactor with a fast neutron spectrum, robust refractory fuel, direct conversion with a gas turbine, and integrated on-site fuel cycle as a promising system for a sustainable energy development. This objective

  14. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  15. Scram device for gas-cooled reactor

    International Nuclear Information System (INIS)

    Murakami, Atsushi; Takahashi, Suehiro.

    1989-01-01

    A scram device for gas-cooled reactors has a hopper disposed below a stand pipe standing upright passing through a reactor container and electromagnets disposed therein. It further comprises neutron absorbing steel balls maintained between the electromagnets and the hopper upon energization of the electromagnets. Upon emergency reactor shutdown, energization for the electromagnets is interrupted to drop the neutron absorption stainless steel balls into the reactor core. It is an object of the present invention to keep the mechanical strength of the electromagnets in a high temperature gas atmosphere and not to reduce the insulation performance. That is, coils for the electromagnets are constituted with a small oxide-insulated metal sheath cable (MI cable). As the feature of the MI cable, it can maintain the mechanical strength even when exposed to high temperature gas coolant and the insulation performance thereof does not reduce by virture of its gas sealing property. Accordingly, a scram device of stable reliability can be obtained. (K.M.)

  16. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

    Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

  17. Status of national gas cooled reactor programmes

    International Nuclear Information System (INIS)

    1991-08-01

    This report has been compiled as a central source of summary-level information on the present status of High Temperature Gas-Cooled Reactor (HTGR) programmes in the world and on future plans for the continued development and deployment of HTGRs. Most of the information concerns the programmes in the United States, Germany, Japan and the Soviet Union, countries that have had large programmes related to HTGR technology for several years. Summary-level information is also provided in the report on HTGR-related activities in several other countries who either have an increasing interest in the technology and/or who are performing some development efforts related to HTGR technology. The report contains a summary-level update on the MAGNOX and AGR programmes. This is the twelfth issue of the document, the first of which was issued in March, 1979. The report has been prepared in the IAEA Nuclear Power Technology Development Section. Figs and tabs

  18. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

    1964-01-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

  19. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-08-01

    On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  20. Gas cooled fast reactor research in Europe

    International Nuclear Information System (INIS)

    Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

    2011-01-01

    Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

  1. Status of and prospects for gas-cooled reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

  2. Development and application of an advanced fuel model for the safety analysis of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Petkevich, P.

    2008-09-01

    Until about the year 2030, current-day nuclear power plants (NPPs) will be replaced by so-called Gen-III or Gen-III+ units, which are mainly based on light water reactor technology. The principal new features are increased safety and improved economical effectiveness. However, these systems use the same fuel forms and are based on the same fuel cycle. Beyond 2030, the interest is likely to shift towards fourth generation NPPs, which offer the possibility of complete fuel cycle closure. Generation-IV reactor concepts include both thermal and fast systems, and involve a wide range of fuel forms and compositions. The present research has been focused on the development of a thermo-mechanical model for the innovative fuel design of the Generation-IV Gas-cooled Fast Reactor (GFR). The principal distinctive feature of the fuel is that the fuel pellets are arranged within plates which enclose an inner honeycomb structure. Apart from the geometry, the usage of new materials is foreseen. Thus, the fuel pellets are of mixed uranium-plutonium carbide, and the cladding is bulk or fiber-reinforced SiC. The setting up of an appropriate materials database was thus the very first task which had to be carried out in the current work. The main purpose of the currently developed model is to provide reliable data, in the context of transient analysis, for the calculation of the principal neutronic feedbacks in the GFR core, viz. the fuel temperature for the Doppler effect and the fuel plate deformation for the axial core expansion effect. None of the available fuel modeling codes is suitable for a realistic simulation of the GFR fuel, as the inner honeycomb structure cannot be explicitly taken into account. The development work has been carried out largely in the context of PSI’s generic code system for fast reactor safety analysis, FAST. Thereby, it has mainly involved extension of the thermo-mechanical code FRED, developed originally for the modeling of traditional rodded fuel

  3. Advanced IGCC/Hydrogen Gas Turbine Development

    Energy Technology Data Exchange (ETDEWEB)

    York, William [General Electric Company, Schenectady, NY (United States); Hughes, Michael [General Electric Company, Schenectady, NY (United States); Berry, Jonathan [General Electric Company, Schenectady, NY (United States); Russell, Tamara [General Electric Company, Schenectady, NY (United States); Lau, Y. C. [General Electric Company, Schenectady, NY (United States); Liu, Shan [General Electric Company, Schenectady, NY (United States); Arnett, Michael [General Electric Company, Schenectady, NY (United States); Peck, Arthur [General Electric Company, Schenectady, NY (United States); Tralshawala, Nilesh [General Electric Company, Schenectady, NY (United States); Weber, Joseph [General Electric Company, Schenectady, NY (United States); Benjamin, Marc [General Electric Company, Schenectady, NY (United States); Iduate, Michelle [General Electric Company, Schenectady, NY (United States); Kittleson, Jacob [General Electric Company, Schenectady, NY (United States); Garcia-Crespo, Andres [General Electric Company, Schenectady, NY (United States); Delvaux, John [General Electric Company, Schenectady, NY (United States); Casanova, Fernando [General Electric Company, Schenectady, NY (United States); Lacy, Ben [General Electric Company, Schenectady, NY (United States); Brzek, Brian [General Electric Company, Schenectady, NY (United States); Wolfe, Chris [General Electric Company, Schenectady, NY (United States); Palafox, Pepe [General Electric Company, Schenectady, NY (United States); Ding, Ben [General Electric Company, Schenectady, NY (United States); Badding, Bruce [General Electric Company, Schenectady, NY (United States); McDuffie, Dwayne [General Electric Company, Schenectady, NY (United States); Zemsky, Christine [General Electric Company, Schenectady, NY (United States)

    2015-07-30

    The objective of this program was to develop the technologies required for a fuel flexible (coal derived hydrogen or syngas) gas turbine for IGCC that met DOE turbine performance goals. The overall DOE Advanced Power System goal was to conduct the research and development (R&D) necessary to produce coal-based IGCC power systems with high efficiency, near-zero emissions, and competitive capital cost. To meet this goal, the DOE Fossil Energy Turbine Program had as an interim objective of 2 to 3 percentage points improvement in combined cycle (CC) efficiency. The final goal is 3 to 5 percentage points improvement in CC efficiency above the state of the art for CC turbines in IGCC applications at the time the program started. The efficiency goals were for NOx emissions of less than 2 ppm NOx (@15 % O2). As a result of the technologies developed under this program, the DOE goals were exceeded with a projected 8 point efficiency improvement. In addition, a new combustion technology was conceived of and developed to overcome the challenges of burning hydrogen and achieving the DOE’s NOx goal. This report also covers the developments under the ARRA-funded portion of the program that include gas turbine technology advancements for improvement in the efficiency, emissions, and cost performance of gas turbines for industrial applications with carbon capture and sequestration. Example applications could be cement plants, chemical plants, refineries, steel and aluminum plants, manufacturing facilities, etc. The DOE’s goal for more than 5 percentage point improvement in efficiency was met with cycle analyses performed for representative IGCC Steel Mill and IGCC Refinery applications. Technologies were developed in this program under the following areas: combustion, larger latter stage buckets, CMC and EBC, advanced materials and coatings, advanced configurations to reduce cooling, sealing and rotor purge flows, turbine aerodynamics, advanced sensors, advancements in first

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-03-01

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

  5. Gas Mixtures for Welding with Micro-Jet Cooling

    Directory of Open Access Journals (Sweden)

    Węgrzyn T.

    2015-04-01

    Full Text Available Welding with micro-jet cooling after was tested only for MIG and MAG processes. For micro-jet gases was tested only argon, helium and nitrogen. A paper presents a piece of information about gas mixtures for micro-jet cooling after in welding. There are put down information about gas mixtures that could be chosen both for MAG welding and for micro-jet process. There were given main information about influence of various micro-jet gas mixtures on metallographic structure of steel welds. Mechanical properties of weld was presented in terms of various gas mixtures selection for micro-jet cooling.

  6. ADVANCED GAS TURBINE SYSTEMS RESEARCH

    Energy Technology Data Exchange (ETDEWEB)

    Unknown

    2002-04-01

    The activities of the Advanced Gas Turbine Systems Research (AGTSR) program for this reporting period are described in this quarterly report. The report is divided into discussions of Membership, Administration, Technology Transfer (Workshop/Education), Research and Miscellaneous Related Activity. Items worthy of note are presented in extended bullet format following the appropriate heading.

  7. ADVANCED GAS TURBINE SYSTEMS RESEARCH

    Energy Technology Data Exchange (ETDEWEB)

    Unknown

    2002-02-01

    The activities of the Advanced Gas Turbine Systems Research (AGTSR) program for this reporting period are described in this quarterly report. The report is divided into discussions of Membership, Administration, Technology Transfer (Workshop/Education), Research and Miscellaneous Related Activity. Items worthy of note are presented in extended bullet format following the appropriate heading.

  8. Buffer gas cooling and mixture analysis

    Science.gov (United States)

    Patterson, David S.; Doyle, John M.

    2018-03-06

    An apparatus for spectroscopy of a gas mixture is described. Such an apparatus includes a gas mixing system configured to mix a hot analyte gas that includes at least one analyte species in a gas phase into a cold buffer gas, thereby forming a supersaturated mixture to be provided for spectroscopic analysis.

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

    International Nuclear Information System (INIS)

    1980-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

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

  11. Cooling performance of helium-gas/water coolers in HENDEL

    International Nuclear Information System (INIS)

    Inagaki, Yoshiyuki; Takada, Shoji; Hayashi, Haruyoshi; Kobayashi, Toshiaki; Ohta, Yukimaru; Shimomura, Hiroaki; Miyamoto, Yoshiaki

    1994-01-01

    The helium engineering demonstration loop (HENDEL) has four helium-gas/water coolers where the cooling water flows in the tubes and helium gas on the shell side. Their cooling performance was studied using the operational data from 1982 to 1991. The heat transfer of helium gas on the shell was obtained for segmental and step-up baffle type coolers. Also, the change with operation time was investigated. The cooling performance was lowered by the graphite powder released from the graphite components for several thousand hours and thereafter recovered because the graphite powder from the components was reduced and the powder in the cooler shell was blown off during the operation. (orig.)

  12. Exergy analysis of a gas-hydrate cool storage system

    International Nuclear Information System (INIS)

    Bi, Yuehong; Liu, Xiao; Jiang, Minghe

    2014-01-01

    Based on exergy analysis of charging and discharging processes in a gas-hydrate cool storage system, the formulas for exergy efficiency at the sensible heat transfer stage and the phase change stage corresponding to gas-hydrate charging and discharging processes are obtained. Furthermore, the overall exergy efficiency expressions of charging, discharging processes and the thermodynamic cycle of the gas-hydrate cool storage system are obtained. By using the above expressions, the effects of number of transfer units, the inlet temperatures of the cooling medium and the heating medium on exergy efficiencies of the gas-hydrate cool storage system are emphatically analyzed. The research results can be directly used to evaluate the performance of gas-hydrate cool storage systems and design more efficient energy systems by reducing the sources of inefficiency in gas-hydrate cool storage systems. - Highlights: • Formulas for exergy efficiency at four stages are obtained. • Exergy efficiency expressions of two processes and one cycle are obtained. • Three mainly influencing factors on exergy efficiencies are analyzed. • With increasing the inlet temperature of cooling medium, exergy efficiency increases. • With decreasing the inlet temperature of heating medium, exergy efficiency increases

  13. Gas cooled fast breeder reactors using mixed carbide fuel

    International Nuclear Information System (INIS)

    Kypreos, S.

    1976-09-01

    The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

  14. Advanced Turbine Blade Cooling Techniques, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Gas turbine engine technology is constantly challenged to operate at higher combustor outlet temperatures. In a modern gas turbine engine, these temperatures can...

  15. Recent advances in cooled-semen technology.

    Science.gov (United States)

    Aurich, Christine

    2008-09-01

    The majority of horse registries approve the use of artificial insemination, and horse breeding has widely taken benefit from the use of cooled-stored semen. New insights into cooled-semen technology open possibilities to reduce problems such as impaired semen quality after cooled-storage in individual stallions. The stallion itself has major impacts on quality and fertility of cooled-stored semen. Dietary supplementation of antioxidants and polyunsaturated fatty acids improves semen quality in a variety of species, but only few studies on this topic exist in the horse. Proper semen collection and handling is the main key to the maintenance of semen quality during cooled-storage. Semen collection should be achieved by minimal sexual stimulation with a single mount; this results in high sperm concentration, low content of seminal plasma and minimal contamination with bacteria. Milk-based semen extenders are most popular for semen processing and storage. The development of more defined extenders containing only the beneficial milk ingredients has made extender quality more constant and reliable. Semen is often centrifuged to decrease the seminal plasma content. Centrifugation results in a recovery rate of only 75% of spermatozoa in the semen pellet. Recovery rates after centrifugation may be improved with use of a "cushion technique" allowing higher centrifugation force and duration. However, this is not routinely used in cooled-semen technology. After slow-cooling, semen-storage and shipping is best performed at 5 degrees C, maintaining semen motility, membrane integrity and DNA integrity for up to 40 h after collection. Shipping containers created from Styrofoam boxes provide maintenance of semen quality at low cost.

  16. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

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

    Energy Technology Data Exchange (ETDEWEB)

    Collins, J.L.

    2004-12-02

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

  18. Gas turbine bucket with impingement cooled platform

    Science.gov (United States)

    Jones, Raphael Durand

    2002-01-01

    In a turbine bucket having an airfoil portion and a root portion, with a substantially planar platform at an interface between the airfoil portion and root portion, a platform cooling arrangement including at least one bore in the root portion and at least one impingement cooling tube seated in the bore, the tube extending beyond the bore with an outlet in close proximity to a targeted area on an underside of the platform.

  19. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    International Nuclear Information System (INIS)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

  20. Overview of gas cooled reactors' applications with CATHARE

    International Nuclear Information System (INIS)

    Genevieve Geffraye; Fabrice Bentivoglio; Anne Messie; Alain Ruby; Manuel Saez; Nicolas Tauveron; Ola Widlund

    2005-01-01

    Full text of publication follows: For about four years, CEA has launched feasibility studies of future nuclear advanced systems in a consistent series of Gas Cooled Reactors (GCR) ranging from thermal reactors, as the Very High Temperature Reactor (VHTR) for the mid term, to fast reactors (GFR) for the long term. Thermal hydraulic performances are a key issue for the core design, the evaluation of the thermal stresses on the structures and the decay heat removal systems. This analysis requires a 1D code able to simulate the whole reactor, including the core, the vessel, the piping and the components (turbine, compressors, heat exchangers). CATHARE is the reference code developed and extensively validated in collaboration between CEA, EDF, IRSN and FRAMATOME-ANP for the French Pressurized Water Reactors. CATHARE has the capabilities to model a Gas Cooled Reactor using standard 0D and 1D modules with some adaptations to treat the specificities of the GCR designs. In this paper, the different adaptations are presented and discussed. The direct coupling of a Gas Cooled Reactor with a closed gas-turbine cycle leads to a specific dynamic plant behaviour and a specific turbomachinery module has been developed. The thermal reactors' core consists of hexagonal graphite blocks with an annular-fueled region surrounded by reflectors and a special attention is paid on the thermal modeling of such a core leading to a quasi-2D thermal description. First designs of the VHTR are proposed and are based on an indirect cycle concept with a primary circuit, cooled by helium, and containing the core and a circulator. The core power is transmitted to the secondary circuit via an intermediate heat exchanger (IHX). The secondary circuit contains a turbine and a compressor coupled on a single shaft. It uses a mixture of helium and nitrogen, in order to benefit from both the favourable thermal properties of helium for the heat exchanger, and from existing experience of turbomachines using

  1. Technology development for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Homan, F.J.; Turner, R.F.

    1989-01-01

    In the USA the Modular High-Temperature Gas-Cooled Reactor is in an advanced stage of design. The related HTGR program areas, the approaches to these programs along with sample results and a description of how these data are used are highlighted in the paper. (author). Figs and tabs

  2. Gas Mixtures for Welding with Micro-Jet Cooling

    OpenAIRE

    Węgrzyn T.

    2015-01-01

    Welding with micro-jet cooling after was tested only for MIG and MAG processes. For micro-jet gases was tested only argon, helium and nitrogen. A paper presents a piece of information about gas mixtures for micro-jet cooling after in welding. There are put down information about gas mixtures that could be chosen both for MAG welding and for micro-jet process. There were given main information about influence of various micro-jet gas mixtures on metallographic structure of steel welds. Mechani...

  3. R + D work on gas-cooled breeder development

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Jacobs, G.; Meyer, L.; Rehme, K.; Schumacher, G.; Wilhelm, D.

    1978-01-01

    The development work for the gas-cooled breeder in the Karlsruhe Nuclear Research Center may be assigned to two different groups: a) Investigations on fuel elements. b) Studies concerning the safety of gas-cooled fast breeder reactors. To the first group there belongs the work related to the: - heat transfer between fuel elements and coolant gas, - influence of increased content of water vapor in helium or the fuel rods. The second group concerns: - establishing a computer code for transient calculations in the primary and secondary circuit of a gas-cooled fast breeder reactor, - steam reactivity coefficients, - the core destruction phase of hypothetical accidents, - the core-catcher using borax. (orig./RW) [de

  4. Utility industry evaluation of the Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Burstein, S.; Bitel, J.S.; Tramm, T.R.; High, M.D.; Neils, G.H.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the Modular High Temperature Gas-Cooled Reactor plant design, a current design created by an industrial team led by General Atomics under Department of Energy sponsorship and with support provided by utilities through Gas-Cooled Reactor Associates. The utility industry team concluded that the plant design should be considered a viable application of an advanced nuclear concept and deserves continuing development. Specific comments and recommendations are provided as a contribution toward improving a very promising plant design. 2 refs

  5. Removal of tritium from gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1976-01-01

    Tritium contained in the coolant gas in the primary circuit of a gas cooled nuclear reactor together with further tritium adsorbed on the graphite used as a moderator for the reactor is removed by introducing hydrogen or a hydrogen-containing compound, for example methane or ammonia, into the coolant gas. The addition of the hydrogen or hydrogen-containing compound to the coolant gas causes the adsorbed tritium to be released into the coolant gas and the tritium is then removed from the coolant gas by passing the mixture of coolant gas and hydrogen or hydrogen-containing compound through a gas purification plant before recirculating the coolant gas through the reactor. 14 claims, 1 drawing figure

  6. The Advanced Gas Centrifuge program

    International Nuclear Information System (INIS)

    Riepe, R.

    1984-01-01

    Although the gas centrifuge process for uranium enrichment is often referred to as a ''new technology,'' it has been under development for approximately 25 years to bring it to its current state of deployment. Centrifuges are now being installed in a new gas centrifuge enrichment plant (GCEP) at Portsmouth, Ohio. The objective of this new plant was to provide additional U.S. uranium enrichment capacity at a production cost comparable to the U.S. diffusion process but requiring much less power per separative work unit (SWU) produced. The current, commercial scale centrifuge technology being installed meets that objective. The objective for new U.S. enrichment capacity has changed. The objective is not to provide more SWUs but to provide cheaper SWUs. The objective is to make the U.S. uranium enrichment enterprise competitive on the international market. Where the U.S. at one time supplied virtually all of the free world SWU demand, the U.S. market share has now dropped to approximately 35% of the foreign free world market. The Advanced Gas Centrifuge (AGC) program provides an avenue for making the U.S. the economically attractive, reliable enrichment supplier

  7. IAEA high temperature gas-cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2000-01-01

    The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

  8. Interotex-innovative gas equipment for heating and cooling

    Energy Technology Data Exchange (ETDEWEB)

    Winnington, T.L. [Interotex Ltd. (United Kingdom); Moore, N. [British Gas plc (United Kingdom); Valle, F.; Sanz, J. I. [Gas Natural SDG S.A. (Spain); Chavarri, J.M. [Fagor Electrodomesticos S. Coop. (Spain); Uselton, R. [Lennox Industries Inc. (United States)

    1997-10-01

    Conventionally, cooling technology for the residential market is provided by electrically driven vapour re-compression systems. But lately, due to the Montreal Protocol - restricting the utilisation of ozone depleting substances - and to the high peak demand in electricity, created by electrical air conditioning systems, there is a commercial opportunity for gas fired air conditioning appliances. This paper describes the development programme for a radical new absorption technology, from the theoretical studies, through the experimental programme, to the building, commissioning and installation of demonstration machines. It also includes an analysis of the world-wide residential cooling market and the opportunities available to manufacturers and gas utilities to introduce new gas heating and cooling technology, capable of competing effectively with electrical systems. (au)

  9. Design requirements, operation and maintenance of gas-cooled reactors

    International Nuclear Information System (INIS)

    1989-06-01

    At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

  10. Testing aspects of advanced coherent electron cooling technique

    Energy Technology Data Exchange (ETDEWEB)

    Litvinenko, V.; Jing, Y.; Pinayev, I.; Wang, G.; Samulyak, R.; Ratner, D.

    2015-05-03

    An advanced version of the Coherent-electron Cooling (CeC) based on the micro-bunching instability was proposed. This approach promises significant increase in the bandwidth of the CeC system and, therefore, significant shortening of cooling time in high-energy hadron colliders. In this paper we present our plans of simulating and testing the key aspects of this proposed technique using the set-up of the coherent-electron-cooling proof-of-principle experiment at BNL.

  11. Evaluation of advanced cooling therapy's esophageal cooling device for core temperature control.

    Science.gov (United States)

    Naiman, Melissa; Shanley, Patrick; Garrett, Frank; Kulstad, Erik

    2016-05-01

    Managing core temperature is critical to patient outcomes in a wide range of clinical scenarios. Previous devices designed to perform temperature management required a trade-off between invasiveness and temperature modulation efficiency. The Esophageal Cooling Device, made by Advanced Cooling Therapy (Chicago, IL), was developed to optimize warming and cooling efficiency through an easy and low risk procedure that leverages heat transfer through convection and conduction. Clinical data from cardiac arrest, fever, and critical burn patients indicate that the Esophageal Cooling Device performs very well both in terms of temperature modulation (cooling rates of approximately 1.3°C/hour, warming of up to 0.5°C/hour) and maintaining temperature stability (variation around goal temperature ± 0.3°C). Physicians have reported that device performance is comparable to the performance of intravascular temperature management techniques and superior to the performance of surface devices, while avoiding the downsides associated with both.

  12. Emergency cooling method and system for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1982-01-01

    For emergency cooling of gas-cooled fast breeder reactors (GSB), which have a core consisting of a fission zone and a breeding zone, water is sprayed out of nozzles on to the core from above in the case of an incident. The water which is not treated with boron is taken out of a reservoir in the form of a storage tank in such a maximum quantity that the cooling water gathering in the space below the core rises at most up to the lower edge of the fission zone. (orig./GL) [de

  13. Heat removal in gas-cooled fuel rod clusters

    International Nuclear Information System (INIS)

    Rehme, K.

    1975-01-01

    For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

  14. Gas-cooled fast reactor safety

    International Nuclear Information System (INIS)

    Rickard, C.L.; Simon, R.H.; Buttemer, D.R.

    1977-01-01

    Initial conceptual design work on the GCFR began in the USA in the early 1960s and since the later 1960s has proceeded with considerable international cooperation. A 300 MWe GCFR demonstration plant employing three main cooling loops is currently being developed at General Atomic. A major preapplication licensing review of this demonstration plant was initiated in 1971 leading in 1974 to publication of a Safety Evaluation Report by the USAEC Directorate of Licensing. The preapplication review is continuing by addressing areas of concern identified in this report such that a major part of the work necessary to support the actual licensing of a GCFR demonstration plant has been established. The safety performance of the GCFR demonstration plant is based upon its inherent safety characteristics among which are the single phase and chemically inert coolant which is not activated and has a low reactivity worth, the negative core power and temperature reactivity coefficients and the small and negative steam reactivity worth. Recent studies of larger core designs indicate that as the reactor size increases central fuel, clad and coolant reactivity worths decrease and the Doppler coefficient becomes more negative. These inherent safety characteristics are complemented by safety design features such as enclosing the entire primary coolant system within a prestressed concrete pressure vessel (PCRV), providing two independent and diverse shutdown systems and residual heat removal (RHR) systems, limiting the worth of control rods to less than $1, employing pressure-equalized fuel rods, a core supported rigidly at its upper end and otherwise unrestrained and coolant downflow within the core to enhance debris removal should local melting occur. The structurally redundant PCRV design allows the potential depressurization leak area to be controlled and, since the PCRV is located within a containment building, coolant is present even after a depressurization accident and each RHR

  15. Cooling of gas turbines IX : cooling effects from use of ceramic coatings on water-cooled turbine blades

    Science.gov (United States)

    Brown, W Byron; Livingood, John N B

    1948-01-01

    The hottest part of a turbine blade is likely to be the trailing portion. When the blades are cooled and when water is used as the coolant, the cooling passages are placed as close as possible to the trailing edge in order to cool this portion. In some cases, however, the trailing portion of the blade is so narrow, for aerodynamic reasons, that water passages cannot be located very near the trailing edge. Because ceramic coatings offer the possibility of protection for the trailing part of such narrow blades, a theoretical study has been made of the cooling effect of a ceramic coating on: (1) the blade-metal temperature when the gas temperature is unchanged, and (2) the gas temperature when the metal temperature is unchanged. Comparison is also made between the changes in the blade or gas temperatures produced by ceramic coatings and the changes produced by moving the cooling passages nearer the trailing edge. This comparison was made to provide a standard for evaluating the gains obtainable with ceramic coatings as compared to those obtainable by constructing the turbine blade in such a manner that water passages could be located very near the trailing edge.

  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. Superconducting cable cooling system by helium gas at two pressures

    International Nuclear Information System (INIS)

    Dean, J.W.

    1977-01-01

    Thermally contacting, oppositely streaming, cryogenic fluid streams in the same enclosure in a closed cycle changes the fluid from a cool high pressure helium gas to a cooler reduced pressure helium gas in an expander so as to be at different temperature ranges and pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T 1 . By first circulating the fluid from a refrigerator at one end of the line as a cool gas at a temperature range T 2 to T 3 in the go leg, then circulating the gas through an expander at the other end of the line where the gas becomes a cooler gas at a reduced pressure and at a reduced temperature T 4 and finally by circulating the cooler gas back again to the refrigerator in a return leg at a temperature range T 4 to T 5 , while in thermal contact with the gas in the go leg, and in the same enclosure therewith for compression into a higher pressure gas at T 2 in a closed cycle, where T 2 greater than T 3 and T 5 greater than T 4 , the fluid leaves the enclosure in the go leg as a gas at its coldest point in the go leg, and the temperature distribution is such that the line temperature decreases along its length from the refrigerator due to the cooling from the gas in the return leg

  18. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

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

  19. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

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

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

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

  1. Periodic reviews of structural integrity of gas-cooled reactors

    International Nuclear Information System (INIS)

    Banks, P.J.; Stokoe, T.Y.; Thomas, D.L.

    1995-01-01

    Nuclear Electric operates 12 gas-cooled reactor power stations which have been in service for between 5 and 30 years. Periodically, comprehensive reviews of the safety cases are carried out for each station. The approach followed in these reviews in respect of structural integrity is outlined with the use of illustrative examples. (author)

  2. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    1985-12-01

    The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

  3. French gas cooled reactor experience with moisture ingress

    International Nuclear Information System (INIS)

    Bastien, D.; Brie, M.

    1995-01-01

    During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

  4. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

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

    1980-03-01

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

  5. Detection of fuel element failures at the gas cooled reactors

    International Nuclear Information System (INIS)

    Bremsak, P.; Ranzel, S.; Ogrin, R.; Pisanski, C.

    1964-01-01

    This paper describes the system for measuring the concentration of released Kr and Xe in the cooling gas. The system developed in the Jozef Stefan Institute is efficient and suitable for application The method is based on electrostatic collection of daughter elements from radioactive decay of Xe and Kr

  6. IAEA activities in gas-cooled reactor technology development

    International Nuclear Information System (INIS)

    Cleveland, J.; Kupitz, J.

    1992-01-01

    The International Atomic Energy Agency (IAEA) has the charter to ''foster the exchange of scientific and technical information'', and ''encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world''. This paper describes the Agency's activities in Gas-cooled Reactor (GCR) technology development

  7. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    International Nuclear Information System (INIS)

    1988-01-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

  8. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-07-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics.

  9. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  10. Heat transport and afterheat removal for gas cooled reactors under accident conditions

    International Nuclear Information System (INIS)

    2001-01-01

    The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

  11. Fuel assembly for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.

    1976-01-01

    A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

  12. Long-term prospects for the gas-cooled reactor

    International Nuclear Information System (INIS)

    Tan, W.P.S.

    1982-01-01

    Towards the second half of a fifty-year time span the market for gas-cooled reactors as sources of high temperature process heat and as highly fuel efficient electricity producers should be reasonably bright, given a fair degree of technological maturity and consequent realisation of inherent economic advantages. Declining fossil resources and increasing prices, initially in oil and gas later in open-cast coal, provide the economic impetus towards substitution of nuclear for coal heat, not only in the generally accepted processes of coal conversion and steel-making but also for oil shale pyrolysis and electrothermal aluminium smelting. Around 2010, if not sooner, the need for uranium conservation should allow the market penetration of breeders and thorium-cycle reactors for which gas cooling has a potential techno-economic edge. (author)

  13. Long-term prospects for the gas-cooled reactor

    International Nuclear Information System (INIS)

    Tan, W.P.S.

    1983-01-01

    Towards the second half of a 50-year time span the market for gas-cooled reactors as sources of high-temperature process heat and as highly fuel-efficient electricity producers should be reasonably bright, given a fair degree of technological maturity and consequent realization of inherent economic advantages. Declining fossil resources and increasing prices, initially in oil and gas, later in open-cast coal, provide the economic impetus towards substitution of nuclear for coal heat, not only in the generally accepted processes of coal conversion and steel making but also for oil shale pyrolysis and electrothermal aluminium smelting. Around 2010, if not sooner, the need for uranium conservation should allow the market penetration of breeders and thorium-cycle reactors for which gas cooling has a potential techno-economic edge. (author)

  14. Corrosion behaviour of high temperature alloys in the cooling gas of high temperature reactors

    International Nuclear Information System (INIS)

    Quadakkers, W.J.; Schuster, H.

    1989-01-01

    The reactive impurities in the primary cooling helium of advanced high temperature gas cooled reactors (HTGR) can cause oxidation, carburization or decarburization of the heat exchanging metallic components. By studies of the fundamental aspects of the corrosion mechanisms it became possible to define operating conditions under which the metallic construction materials show, from the viewpoint of technical application, acceptable corrosion behaviour. By extensive test programmes with exposure times of up to 30,000 hours, a data base has been obtained which allows a reliable extrapolation of the corrosion effects up to the envisaged service lives of the heat exchanging components. (author). 6 refs, 7 figs

  15. Gas-Cooled Reactors: the importance of their development

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1978-01-01

    Gas-Cooled Reactors are considered to have a significant future impact on the application of fission energy. The specific types are the steam-cycle High-Temperature Gas-Cooled Reactor, the Gas-Cooled Fast Breeder Reactor, the gas-turbine HTGR, and the Very High-Temperature Process Heat Reactor. The importance of developing the above systems is discussed relative to alternative fission power systems involving Light Water Reactors, Heavy Water Reactors, Spectral Shift Controlled Reactors, and Liquid-Metal-Cooled Fast Breeder Reactors. A primary advantage of developing GCRs as a class lies in the technology and cost interrelations, permitting cost-effective development of systems having diverse applications. Further, HTGR-type systems have highly proliferation-resistant characteristics and very attractive safety features. Finally, such systems and GCFRs are mutally complementary. Overall, GCRs provide interrelated systems that serve different purposes and needs; their development can proceed in stages that provide early benefits while contributing to future needs. It is concluded that the long-term importance of the various GCRs is as follows: HTGR, providing a technology for economic GCFRs and HTGR-GTs, while providing a proliferation-resistant reactor system having early economic and fuel utilization benefits; GCFR, providing relatively low cost fissile fuel and reducing overall separative work needs at capital costs lower than those for LMFBRs; HTGR-GT (in combination with a bottoming cycle), providing a very high thermal efficiency system having low capital costs and improved fuel utilization and technology pertinent to VHTRs; HTGR-GT, providing a power system well suited for dry cooling conditions for low-temperature process heat needs; and VHTR, providing a high-temperature heat source for hydrogen production processes

  16. High-temperature gas-cooled reactors and process heat

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1980-01-01

    High-Temperature Gas-Cooled Reactors (HTGRs) are fueled with ceramic-coated microspheres of uranium and thorium oxides/carbides embedded in graphite blocks which are cooled with helium. Promising areas of HTGR application are in cogeneration, energy transport using Heat Transfer Salt, recovery of oils from oil shale, steam reforming of methane for chemical production, coal gasification, and in energy transfer using chemical heat jpipes in the long term. Further, HTGRs could be used as the energy source for hydrogen production through thermochemical water splitting in the long term. The potential market for Process Heat HTGRs is 100-200 large units by about the year 2020

  17. Absorber rod driving into a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Elter, C.; Schmitt, H.; Schoening, J.

    1987-01-01

    The absorber rod consists of a hollow cylinder which has a layer of absorber material applied on its inside circumferential surface. The absorber rod is held via a guide sleeve, which is supported centrally in a hole in the side reflector. The guidance within the sleeve is provided by flanges on the hollow cylinder. The movement of the hollow cylinder is carried out hydraulically or pneumatically. A flow of cooling gas is used for cooling, which is passed through the inner central areas of the hollow cylinder and the guide sleeve. (DG) [de

  18. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  19. Preliminary design of the cooling system for a gas-cooled, high-fluence fast pulsed reactor (HFFPR)

    International Nuclear Information System (INIS)

    Monteith, H.C.

    1978-10-01

    The High-Fluence Fast Pulsed Reactor (HFFPR) is a research reactor concept currently being evaluated as a source for weapon effects experimentation and advanced reactor safety experiments. One of the designs under consideration is a gas-cooled design for testing large-scale weapon hardware or large bundles of full-length, fast reactor fuel pins. This report describes a conceptual cooling system design for such a reactor. The primary coolant would be helium and the secondary coolant would be water. The size of the helium-to-water heat exchanger and the water-to-water heat exchanger will be on the order of 0.9 metre (3 feet) in diameter and 3 metres (10 feet) in length. Analysis indicates that the entire cooling system will easily fit into the existing Sandia Engineering Reactor Facility (SERF) building. The alloy Incoloy 800H appears to be the best candidate for the tube material in the helium-to-water heat exchanger. Type 316 stainless steel has been recommended for the shell of this heat exchanger. Estimates place the cost of the helium-to-water heat exchanger at approximately $100,000, the water-to-water heat exchanger at approximately $25,000, and the helium pump at approximately $450,000. The overall cost of the cooling system will approach $2 million

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-09-01

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

  1. IAEA'S study on advanced applications of water cooled nuclear power plants

    International Nuclear Information System (INIS)

    Cleveland, J.; McDonald, A.; Rao, A.; )

    2008-01-01

    About one-fifth of the world's energy consumption is used for electricity generation, with nuclear power contributing approximately 15.2% of this electricity. However; most of the world's energy consumption is for heat and transportation. Nuclear energy has considerable potential to penetrate these energy sectors now served by fossil fuels that are characterized by price volatility and finite supply. Advanced applications of nuclear energy include seawater desalination, district heating, and heat for industrial processes. Nuclear energy also has potential to provide a near-term, greenhouse gas free, source of energy for transportation. These applications rely on a source of heat and electricity. Nuclear energy from water-cooled reactors, of course, is not unique in this sense. Indeed, higher temperature heat can be produced by burning natural gas and coal, or through the use of other nuclear technologies such as gas-cooled or liquid-metal-cooled reactors. Water-cooled reactors, however; are being deployed today while other reactor types have had considerably less operational and regulatory experience and will take still some time to be widely accepted in the market. Both seawater desalination and district heating with nuclear energy are well proven, and new seawater desalination projects using water-cooled reactors will soon be commissioned. Provision of process heat with nuclear energy can result in less dependence on fossil fuels and contribute to reductions of greenhouse gases. Importantly, because nuclear power produces base-load electricity at stable and predictable prices, it provides a greenhouse gas free source of electricity for transportation systems (trains and subways), and for electric and plug-in hybrid vehicles, and in the longer term nuclear energy could produce hydrogen for fuel cell vehicles, as well as for other components of a hydrogen economy. These advanced applications can play an important role in enhancing public acceptance of nuclear

  2. Experimental investigation of temperature rise in bone drilling with cooling: A comparison between modes of without cooling, internal gas cooling, and external liquid cooling.

    Science.gov (United States)

    Shakouri, Ehsan; Haghighi Hassanalideh, Hossein; Gholampour, Seifollah

    2018-01-01

    Bone fracture occurs due to accident, aging, and disease. For the treatment of bone fractures, it is essential that the bones are kept fixed in the right place. In complex fractures, internal fixation or external methods are used to fix the fracture position. In order to immobilize the fracture position and connect the holder equipment to it, bone drilling is required. During the drilling of the bone, the required forces to chip formation could cause an increase in the temperature. If the resulting temperature increases to 47 °C, it causes thermal necrosis of the bone. Thermal necrosis decreases bone strength in the hole and, subsequently, due to incomplete immobilization of bone, fracture repair is not performed correctly. In this study, attempts have been made to compare local temperature increases in different processes of bone drilling. This comparison has been done between drilling without cooling, drilling with gas cooling, and liquid cooling on bovine femur. Drilling tests with gas coolant using direct injection of CO 2 and N 2 gases were carried out by internal coolant drill bit. The results showed that with the use of gas coolant, the elevation of temperature has limited to 6 °C and the thermal necrosis is prevented. Maximum temperature rise reached in drilling without cooling was 56 °C, using gas and liquid coolant, a maximum temperature elevation of 43 °C and 42 °C have been obtained, respectively. This resulted in decreased possibility of thermal necrosis of bone in drilling with gas and liquid cooling. However, the results showed that the values obtained with the drilling method with direct gas cooling are independent of the rotational speed of drill.

  3. Ambient air cooling arrangement having a pre-swirler for gas turbine engine blade cooling

    Science.gov (United States)

    Lee, Ching-Pang; Tham, Kok-Mun; Schroeder, Eric; Meeroff, Jamie; Miller, Jr., Samuel R; Marra, John J

    2015-01-06

    A gas turbine engine including: an ambient-air cooling circuit (10) having a cooling channel (26) disposed in a turbine blade (22) and in fluid communication with a source (12) of ambient air: and an pre-swirler (18), the pre-swirler having: an inner shroud (38); an outer shroud (56); and a plurality of guide vanes (42), each spanning from the inner shroud to the outer shroud. Circumferentially adjacent guide vanes (46, 48) define respective nozzles (44) there between. Forces created by a rotation of the turbine blade motivate ambient air through the cooling circuit. The pre-swirler is configured to impart swirl to ambient air drawn through the nozzles and to direct the swirled ambient air toward a base of the turbine blade. The end walls (50, 54) of the pre-swirler may be contoured.

  4. Power Conversion Study for High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Chang Oh; Richard Moore; Robert Barner

    2005-01-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gas cooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language

  5. Optimal design of gas adsorption refrigerators for cryogenic cooling

    Science.gov (United States)

    Chan, C. K.

    1983-01-01

    The design of gas adsorption refrigerators used for cryogenic cooling in the temperature range of 4K to 120K was examined. The functional relationships among the power requirement for the refrigerator, the system mass, the cycle time and the operating conditions were derived. It was found that the precool temperature, the temperature dependent heat capacities and thermal conductivities, and pressure and temperature variations in the compressors have important impacts on the cooling performance. Optimal designs based on a minimum power criterion were performed for four different gas adsorption refrigerators and a multistage system. It is concluded that the estimates of the power required and the system mass are within manageable limits in various spacecraft environments.

  6. Gas turbine cooling modeling - Thermodynamic analysis and cycle simulations

    Energy Technology Data Exchange (ETDEWEB)

    Jordal, Kristin

    1999-02-01

    Considering that blade and vane cooling are a vital point in the studies of modern gas turbines, there are many ways to include cooling in gas turbine models. Thermodynamic methods for doing this are reviewed in this report, and, based on some of these methods, a number of model requirements are set up and a Cooled Gas Turbine Model (CGTM) for design-point calculations of cooled gas turbines is established. Thereafter, it is shown that it is possible to simulate existing gas turbines with the CGTM. Knowledge of at least one temperature in the hot part of the turbine (TET, TRIT or possibly TIT) is found to be vital for a complete heat balance over the turbine. The losses, which are caused by the mixing of coolant and main flow, are in the CGTM considered through a polytropic efficiency reduction factor S. Through the study of S, it can be demonstrated that there is more to gain from coolant reduction in a small and/or old turbine with poor aerodynamics, than there is to gain in a large, modern turbine, where the losses due to interaction between coolant and main flow are, relatively speaking, small. It is demonstrated, at the design point (TET=1360 deg C, {pi}=20) for the simple-cycle gas turbine, that heat exchanging between coolant and fuel proves to have a large positive impact on cycle efficiency, with an increase of 0.9 percentage points if all of the coolant passes through the heat exchanger. The corresponding improvement for humidified coolant is 0.8 percentage points. A design-point study for the HAT cycle shows that if all of the coolant is extracted after the humidification tower, there is a decrease in coolant requirements of 7.16 percentage points, from 19.58% to 12.52% of the compressed air, and an increase in thermal efficiency of 0.46 percentage points, from 53.46% to 53.92%. Furthermore, it is demonstrated with a TET-parameter variation, that the cooling of a simple-cycle gas turbine with humid air can have a positive effect on thermal efficiency

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

  8. Multidisciplinary design optimization of film-cooled gas turbine blades

    OpenAIRE

    Shashishekara S. Talya; J. N. Rajadas; A. Chattopadhyay

    1999-01-01

    Design optimization of a gas turbine blade geometry for effective film cooling toreduce the blade temperature has been done using a multiobjective optimization formulation. Three optimization formulations have been used. In the first, the average blade temperature is chosen as the objective function to be minimized. An upper bound constraint has been imposed on the maximum blade temperature. In the second, the maximum blade temperature is chosen as the objective function to be minimized with ...

  9. The status of graphite development for gas cooled reactors

    International Nuclear Information System (INIS)

    1993-02-01

    The meeting was convened by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors. It was attended by 61 participants from 6 countries. The meeting covered the following subjects: overview of national programs; design criteria, fracture mechanisms and component test; materials development and properties; non-destructive examination, inspection and surveillance. The participants presented 33 papers on behalf of their countries. A separate abstract was prepared for each of these papers. Refs, figs, tabs, photos and diagrams

  10. Advanced technologies for water cooled reactors 1990. Pt. 1

    International Nuclear Information System (INIS)

    1991-05-01

    The meeting was attended by 20 participants from 12 countries who reviewed and discussed the status and progress of national programmes on advanced water-cooled reactors and recommended to the Scientific Secretary a comprehensive programme for 1991/1992 which would support technology development programmes in IWGATWR Member States. This summary report outlines the activities of IWGATWR since its Second Meeting in June 1988 and main results of the Third Meeting

  11. Advanced technologies for water cooled reactors 1990. Pt. 2

    International Nuclear Information System (INIS)

    1991-05-01

    The main purpose of the meeting was to review and discuss the status of national programmes, the progress achieved since the last meeting held in June 1988 in the field of advanced technologies and design trends for existing and future water cooled reactors. 24 specialists from 14 countries and the IAEA took part in the meeting and 12 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  12. The Formation and Physical Origin of Highly Ionized Cooling Gas

    Energy Technology Data Exchange (ETDEWEB)

    Bordoloi, Rongmon [MIT-Kavli Center for Astrophysics and Space Research, 77 Massachusetts Avenue, Cambridge, MA, 02139 (United States); Wagner, Alexander Y. [University of Tsukuba, Center for Computational Sciences, Tennodai 1-1-1, Tsukuba, Ibaraki (Japan); Heckman, Timothy M.; Norman, Colin A., E-mail: bordoloi@mit.edu, E-mail: bordoloi@mit.edu [Department of Physics and Astronomy, John Hopkins University, 21218, Baltimore, MD (United States)

    2017-10-20

    We present a simple model that explains the origin of warm, diffuse gas seen primarily as highly ionized absorption-line systems in the spectra of background sources. We predict the observed column densities of several highly ionized transitions such as O vi, O vii, Ne viii, N v, and Mg x, and we present a unified comparison of the model predictions with absorption lines seen in the Milky Way disk, Milky Way halo, starburst galaxies, the circumgalactic medium, and the intergalactic medium at low and high redshifts. We show that diffuse gas seen in such diverse environments can be simultaneously explained by a simple model of radiatively cooling gas. We show that most such absorption-line systems are consistent with being collisionally ionized, and we estimate the maximum-likelihood temperature of the gas in each observation. This model satisfactorily explains why O vi is regularly observed around star-forming low- z L* galaxies, and why N v is rarely seen around the same galaxies. We further present some consequences of this model in quantifying the dynamics of the cooling gas around galaxies and predict the shock velocities associated with such flows. A unique strength of this model is that while it has only one free (but physically well-constrained) parameter, it nevertheless successfully reproduces the available data on O vi absorbers in the interstellar, circumgalactic, intragroup, and intergalactic media, as well as the available data on other absorption lines from highly ionized species.

  13. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

  14. Risk Based Inspection of Gas-Cooling Heat Exchanger

    Directory of Open Access Journals (Sweden)

    Dwi Priyanta

    2017-09-01

    Full Text Available On October 2013, Pertamina Hulu Energi Offshore North West Java (PHE – ONWJ platform personnel found 93 leaking tubes locations in the finfan coolers/ gas-cooling heat exchanger. After analysis had been performed, the crack in the tube strongly indicate that stress corrosion cracking was occurred by chloride. Chloride stress corrosion cracking (CLSCC is the cracking occurred by the combined influence of tensile stress and a corrosive environment. CLSCC is the one of the most common reasons why austenitic stainless steel pipework or tube and vessels deteriorate in the chemical processing, petrochemical industries and maritime industries. In this thesis purpose to determine the appropriate inspection planning for two main items (tubes and header box in the gas-cooling heat exchanger using risk based inspection (RBI method. The result, inspection of the tubes must be performed on July 6, 2024 and for the header box inspection must be performed on July 6, 2025. In the end, RBI method can be applicated to gas-cooling heat exchanger. Because, risk on the tubes can be reduced from 4.537 m2/year to 0.453 m2/year. And inspection planning for header box can be reduced from 4.528 m2/year to 0.563 m2/year.

  15. The role of the IAEA in advanced technologies for water-cooled reactors

    International Nuclear Information System (INIS)

    Cleveland, J.

    1996-01-01

    The role of the IAEA in advanced technologies for water-cooled reactors is described, including the following issues: international collaboration ways through international working group activities; IAEA coordinated research programmes; cooperative research in advanced water-cooled reactor technology

  16. Natural convection in closed vertical cylinders with particular reference to gas cooled reactor standpipes

    International Nuclear Information System (INIS)

    Spence, I.D.

    1975-09-01

    The access to the core for fuel assemblies and control rods of the Advanced Gas Cooled Reactor is through the top cap by means of standpipes. The standpipe is essentially a cylindrical, vertical tube with cooled side wall, closed upper end and an orifice at the lower end which is exposed to the hot core fluid. This creates confined natural convection flow in the empty standpipe and this is the subject of this thesis. The investigation is carried out using analytical and experimental methods. For the analytical work, solution of laminar and turbulent flow is attempted using finite-difference computer techniques. The laminar flow performance is evaluated using two different finite-difference procedures, and the results are compared to each other and to existing analytical and experimental results for the open thermosyphon with cool inflow and hot sidewall, i.e. the complementary problem to the present one. For turbulent flow a two equation turbulence model is employed which provides transport equations for the kinetic energy of turbulence and its dissipation rate. The experimental rig is a full scale replica of the Advanced Gas Cooled Reactor control rod mechanism standpipe. Carbon dioxide and helium are used as the working fluids for the series of tests. (author)

  17. Advances in gas avalanche photomultipliers

    CERN Document Server

    Breskin, Amos; Buzulutskov, A F; Chechik, R; Garty, E; Shefer, G; Singh, B K

    2000-01-01

    Gas avalanche detectors, combining solid photocathodes with fast electron multipliers, provide an attractive solution for photon localization over very large sensitive areas and under high illumination flux. They offer single-photon sensitivity and the possibility of operation under very intense magnetic fields. We discuss the principal factors governing the operation of gas avalanche photomultipliers. We summarize the recent progress made in alkali-halide and CVD-diamond UV-photocathodes, capable of operation under gas multiplication, and novel thin-film protected alkali-antimonide photocathodes, providing, for the first time, the possibility of operating gas photomultipliers in the visible range. Electron multipliers, adequate for these photon detectors, are proposed and some applications are briefly discussed.

  18. The modular high-temperature gas-cooled reactor (MHTGR) in the US

    International Nuclear Information System (INIS)

    Neylan, A.J.; Graf, D.F.; Millunzi, A.C.

    1987-01-01

    GA Technologies Inc. and other U.S. corporations, in a cooperative program with the U.S. Department of Energy, is developing a Modular High-Temperature Gas-Cooled Reactor (MHTGR) that will provide highly reliable, economic, nuclear power. The MHTGR system assures maximum safety to the public, the owner/operator, and the environment. The MHTGR is being designed to meet and exceed rigorous requirements established by the user industry for availability, operation and maintenance, plant investment protection, safety and licensing, siting flexibility and economics. The plant will be equally attractive for deployment and operation in the U.S., other major industrialized nations including Korea, Japan, and the Republic of China, as well as the developing nations. The High-Temperature Gas-Cooled Reactor (HTGR) is an advanced, third generation nuclear power system which incorporates distinctive technical features, including the use of pressurized helium as a coolant, graphite as the moderator and core structural material, and fuel in the form of ceramic coated uranium particles. The modular HTGR builds upon generic gas-cooled reactor experience and specific HTGR programs and projects. The MHTGR offers unique technological features and the opportunity for the cooperative international development of an advanced energy system that will help assure adaquate world energy resources for the future. Such international joint venturing of energy development can offer significant benefits to participating industries and governments and also provides a long term solution to the complex problems of the international balance of payments

  19. Advanced ultrasonic technology for natural gas measurement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    In recent years, due to rising environmental and safety concerns, increasing commodity prices, and operational inefficiencies, a paradigm shift has been taking place with respect to gas measurement. The price of natural gas depends on the location, time of the year, and type of consumer. There is wide uncertainty associated with an orifice meter. This paper presents the use of advanced ultrasonic technology for the measurement of natural gas. For many years, multi-path ultrasonic meters with intelligent sensor technology have been used for gas measurement. This paper gives the various applications of ultrasonic technology along with their advantages and a draws a comparison with orifice meters. From the study it can be concluded that extensive advances in the use of ultrasonic technology for gas measurement have widened the areas of application and that varying frequencies combined with sealed transducer designs make it possible to measure atmospheric and sour gas in custody transfer process control and flaring accurately.

  20. Advancing agricultural greenhouse gas quantification*

    Science.gov (United States)

    Olander, Lydia; Wollenberg, Eva; Tubiello, Francesco; Herold, Martin

    2013-03-01

    given to methods appropriate to low-income countries, where strategies are needed for getting robust data with extremely limited resources in order to support national mitigation planning within widely accepted standards and thus provide access to essential international support, including climate funding. Managing agricultural emissions needs to occur in tandem with managing for agricultural productivity, resilience to climate change, and ecosystem impacts. Management decisions and priorities will require measures and information that identify GHG efficiencies in production and reduce inputs without reducing yields, while addressing climate resilience and maintaining other essential environmental services, such as water quality and support for pollinators. Another set of papers in this issue considers the critical synergies and tradeoffs possible between these multiple objectives of mitigation, resilience, and production efficiency to help us understand how we need to tackle these in our quantification systems. Significant capacity to quantify greenhouse gases is already built, and with some near-term strategic investment, could become an increasingly robust and useful tool for planning and development in the agricultural sector around the world. Acknowledgments The Climate Change Agriculture and Food Security Program of the Consultative Group on International Agricultural Research, the Technical Working Group on Agricultural Greenhouse Gases (T-AGG) at Duke University's Nicholas Institute for Environmental Policy Solutions, and the United Nations Food and Agriculture Organization (FAO) have come together to guide the development of this focus issue and associated activities and papers, given their common desire to improve our understanding of the state of agricultural greenhouse gas (GHG) quantification and to advance ideas for building data and methods that will help mitigation policy and programs move forward around the world. We thank the David and Lucile Packard

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

    International Nuclear Information System (INIS)

    Lap-Yan, C.; Wie, T. Y. C.

    2009-01-01

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

  2. Heat transfer problems in gas-cooled solid blankets

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  3. Verification of Thermal Models of Internally Cooled Gas Turbine Blades

    Directory of Open Access Journals (Sweden)

    Igor Shevchenko

    2018-01-01

    Full Text Available Numerical simulation of temperature field of cooled turbine blades is a required element of gas turbine engine design process. The verification is usually performed on the basis of results of test of full-size blade prototype on a gas-dynamic test bench. A method of calorimetric measurement in a molten metal thermostat for verification of a thermal model of cooled blade is proposed in this paper. The method allows obtaining local values of heat flux in each point of blade surface within a single experiment. The error of determination of local heat transfer coefficients using this method does not exceed 8% for blades with radial channels. An important feature of the method is that the heat load remains unchanged during the experiment and the blade outer surface temperature equals zinc melting point. The verification of thermal-hydraulic model of high-pressure turbine blade with cooling allowing asymmetrical heat removal from pressure and suction sides was carried out using the developed method. An analysis of heat transfer coefficients confirmed the high level of heat transfer in the leading edge, whose value is comparable with jet impingement heat transfer. The maximum of the heat transfer coefficients is shifted from the critical point of the leading edge to the pressure side.

  4. Review of RSG-GAS secondary cooling pump performance

    International Nuclear Information System (INIS)

    Marsahala, Y.B.

    1999-01-01

    The control system of RSG-GAS secondary pump is the study for the operation existence of RSG-GAS secondary pump. The research is about characteristic of the secondary pump and its control system. The measuring of characteristic parameter of secondary cooling pump was being done while the pump running. The pump was loading with capacity 1950 m3/hr. with ambient temperature 28.5 oC. The fault effect of public grid (PLN) such as the fluctuation of both voltage and frequency likes voltage drops (dip). Supply block out that effect of the electric motor performances directly will be analyzed. How far those faults will effect the overall performance of secondary cooling system. Analyzing. Will be done according to the control system was installed. Has be done to find the direct effects of the motor performances against the motor rotation fluctuation which run from 1450 rpm to 1475 rpm. The using of start-delta starting method with delay time about 6 seconds, is enough or not to reduce the inrush starting current also analyzed in this paper. From the research can be obtained that in the steady state condition , the electric motor runs with both power and current are still under tolerances permitted. According to the analyzed data above, it will be consider that the control system of secondary pump would be modified or not. Therefore the analyzed data can show the characteristic curve of the secondary cooling system performance

  5. Medium-size high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760 0 C (1400 0 F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

  6. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

    2010-03-01

    At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

  7. Design and development of gas cooled reactors with closed cycle gas turbines. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    Technological advances over the past fifteen years in the design of turbomachinery, recuperators and magnetic bearings provide the potential for a quantum improvement in nuclear power generation economics through the use of the HTGR with a closed cycle gas turbine. Enhanced international co-operation among national gas cooled reactor programmes in these common technology areas could facilitate the development of this nuclear power concept thereby achieving safety, environmental and economic benefits with overall reduced development costs. This TCM and Workshop was convened to provide the opportunity to review and examine the status of design activities and technology development in national HTGR programmes with specific emphasis on the closed cycle gas turbine, and to identify pathways which take advantage of the opportunity for international co-operation in the development of this concept. Refs, figs, tabs.

  8. Design and development of gas cooled reactors with closed cycle gas turbines. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1996-08-01

    Technological advances over the past fifteen years in the design of turbomachinery, recuperators and magnetic bearings provide the potential for a quantum improvement in nuclear power generation economics through the use of the HTGR with a closed cycle gas turbine. Enhanced international co-operation among national gas cooled reactor programmes in these common technology areas could facilitate the development of this nuclear power concept thereby achieving safety, environmental and economic benefits with overall reduced development costs. This TCM and Workshop was convened to provide the opportunity to review and examine the status of design activities and technology development in national HTGR programmes with specific emphasis on the closed cycle gas turbine, and to identify pathways which take advantage of the opportunity for international co-operation in the development of this concept. Refs, figs, tabs

  9. Gas-cooled reactor application for a university campus

    International Nuclear Information System (INIS)

    Colak, Ue.; Kadiroghlu, O.K.; Soekmen, C.N.; Schmitt, H.

    1991-01-01

    Large urban areas with unfavourable topographic and meteorological conditions suffer severe air pollution during the winter months. Use of low grade lignites, imported higher quality coal or imported fuel oil are the sources of air pollution in the form of sulphur dioxide, fly ash and soot. Large housing complexes or old and historical locations within the city are in need of pollution free centralized district heating systems. Natural gas imported from the Soviet Union is a solution for this problem. Lack of gas distribution network for high pressure gas within the city is the main bottle-neck for the heating systems utilizing natural gas. Concern of the safety of flammable high pressure gas circulating within the city is another drawback for the natural gas heating systems. Nuclear district heating is an environmentally viable option worth looking into it. Localized urban nuclear heating is an interesting solution for large urban areas with old and historical character. The results of a feasibility study on the HGR application for the Hacettepe University presented here, summarizes the concept of gas-cooled heating reactors specially designed for urban centers. The inherently safe characteristics of the pebble bed heating reactor makes localized urban nuclear heating a viable alternative to other heat sources. An economical analysis of various heat sources with equal power levels is done for the Beytepe campus of Hacettepe University in Ankara. Under special boundary conditions, the price for heat generation can be much lower for nuclear heating with GHR 20 than for hard coal or fuel oil. It is also possible that if the price escalation rate for natural gas exceeds 3%, then nuclear heating with GHR can be more competitive. It is concluded that the nuclear heating of Beytepe campus with a GHR 20 is feasible and economical. (author) 3 figs., 5 refs

  10. Thermodynamic analysis of turbine blade cooling on the performance of gas turbine cycle

    International Nuclear Information System (INIS)

    Sarabchi, K.; Shokri, M.

    2002-01-01

    Turbine inlet temperature strongly affects gas turbine performance. Today blade cooling technologies facilitate the use of higher inlet temperatures. Of course blade cooling causes some thermodynamic penalties that destroys to some extent the positive effect of higher inlet temperatures. This research aims to model and evaluate the performance of gas turbine cycle with air cooled turbine. In this study internal and transpiration cooling methods has been investigated and the penalties as the result of gas flow friction, cooling air throttling, mixing of cooling air flow with hot gas flow, and irreversible heat transfer have been considered. In addition, it is attempted to consider any factor influencing actual conditions of system in the analysis. It is concluded that penalties due to blade cooling decrease as permissible temperature of the blade surface increases. Also it is observed that transpiration method leads to better performance of gas turbine comparing to internal cooling method

  11. Consideration of emergency source terms for pebble-bed high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Tao, Liu; Jun, Zhao; Jiejuan, Tong; Jianzhu, Cao

    2009-01-01

    Being the last barrier in the nuclear power plant defense-in-depth strategy, emergency planning (EP) is an integrated project. One of the key elements in this process is emergency source terms selection. Emergency Source terms for light water reactor (LWR) nuclear power plant (NPP) have been introduced in many technical documents, and advanced NPP emergency planning is attracting attention recently. Commercial practices of advanced NPP are undergoing in the world, pebble-bed high-temperature gas-cooled reactor (HTGR) power plant is under construction in China which is considered as a representative of advanced NPP. The paper tries to find some pieces of suggestion from our investigation. The discussion of advanced NPP EP will be summarized first, and then the characteristics of pebble-bed HTGR relating to EP will be described. Finally, PSA insights on emergency source terms selection and current pebble-bed HTGR emergency source terms suggestions are proposed

  12. Gas-cooled reactor coolant circulator and blower technology

    International Nuclear Information System (INIS)

    1988-08-01

    In the previous 17 meetings held within the framework of the International Working Group on Gas-Cooled Reactors, a wide variety of topics and components have been addressed, but the San Diego meeting represented the first time that a group of specialists had been convened to discuss circulator and blower related technology. A total of 20 specialists from 6 countries attended the meeting in which 15 technical papers were presented in 5 sessions: circulator operating experience I and II (6 papers); circulator design considerations I and II (6 papers); bearing technology (3 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  13. Gas-cooled reactor power systems for space

    International Nuclear Information System (INIS)

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

  14. Refueling system for the gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1980-05-01

    Criteria specifically related to the handling of Gas-Cooled Fast Breeder Reactor (GCFR) fuel are briefly reviewed, and the most significant requirements with which the refueling system must comply are discussed. Each component of the refueling system is identified, and a functional description of the fuel handling machine is presented. An illustrated operating sequence describing the various functions involved in a typical refueling cycle is presented. The design status of components and subsystems selected for conceptual development is reviewed, and anticipated refueling time frames are given

  15. The UK gas-cooled reactor programme - Progress report 1988

    International Nuclear Information System (INIS)

    Askew, J.R.

    1989-01-01

    This paper summarises key developments during 1988 on the 26 Magnox reactors and 14 AGRs now operating in the UK. Details are given of long-term safety reviews of the Berkeley and Bradwell Magnox stations which resulted in a decision by CEGB to cease generation at Berkeley but to continue operation at Bradwell. The summary of operating experience with the AGRs concentrates on the completion of construction and successful commissioning of the second generation AGRs at Heysham 2 and Torness. An appended article by John Wilson, Deputy Director of the UKAEA's gas-cooled reactor R and D programme, gives details of the aims and achievements of the programme during 1988. (author)

  16. Graphite development for gas-cooled reactors in the USA

    International Nuclear Information System (INIS)

    Burchell, T.D.

    1991-01-01

    This document discusses Modular High-Temperature Gas-Cooled Reactor (MHTGR) graphite activities in the USA which currently include the following research and development tasks: coke examination; effects of irradiation; variability of physical properties (mechanical, thermal-physical, and fracture); fatigue behavior, oxidation behavior; NDE techniques; structural design criteria; and carbon-carbon composite control rod clad materials. These tasks support nuclear grade graphite manufacturing technology including nondestructive examination of billets and components. Moreover, data shall be furnished to support design and licensing of graphite components for the MHTGR

  17. Beam cooling using a gas-filled RFQ ion guide

    CERN Document Server

    Henry, S; De Saint-Simon, M; Jacotin, M; Képinski, J F; Lunney, M D

    1999-01-01

    A radiofrequency quadrupole mass filter is being developed for use as a high-transmission beam cooler by operating it in buffer gas at high pressure. Such a device will increase the sensitivity of on-line experiments that make use of weakly produced radioactive ion beams. We present simulations and some preliminary measurements for a device designed to cool the beam for the MISTRAL RF mass spectrometer on- line at ISOLDE. The work is carried out partly within the frame of the European Community research network: EXOTRAPS. (9 refs).

  18. R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

    2005-01-01

    The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

  19. The cryogenic cooling program at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Rogers, C.S.; Mills, D.M.; Assoufid, L.

    1994-06-01

    This paper describes the experimental and analytical program in cryogenic cooling of high-heat-load optics at the Advanced-Photon Source. A prototype liquid nitrogen pumping system has been procured. This pump provides a variable flow rate of 1 to 10 gpm of pressurized liquid nitrogen and is sized to handle up to 5 kW of optic heat load. Also, a high-vacuum, double-crystal monochromator testing tank has been fabricated. This system will be used to test cryogenic crystals at existing synchrotron sources. A finite element analysis has been performed for a cryogenically cooled Si crystal in the inclined geometry for Undulator A at 100 mA. The inclination angle was 80 degrees. It was set to diffract from the (111) planes at the first harmonic energy of 4.2 keV. The maximum slope error in the diffraction plane was calculated to be about 1 μrad with a peak temperature of 94 K. An analysis has also been performed for a cryogenically-cooled ''thin'' crystal oriented in the Bragg geometry which accepts 87% of the lst harmonic photons at 3.866 keV. The total absorbed power was 131 W at 100 mA current and the peak temperature was 124 K

  20. Commercial sector gas cooling technology frontier and market share analysis

    International Nuclear Information System (INIS)

    Pine, G.D.; Mac Donald, J.M.; McLain, H.A.

    1990-01-01

    This paper describes a method, developed for the Gas Research Institute of the United States, that can assist planning for commercial sector natural gas cooling systems R and D. These systems are higher in first cost than conventional electric chillers. Yet, engine-driven chiller designs exist which are currently competitive in U.S. markets typified by high electricity or demand charges. Section II describes a scenario analysis approach used to develop and test the method. Section III defines the technology frontier, a conceptual tool for identifying new designs with sales potential. Section IV describes a discrete choice method for predicting market shares of technologies with sales potential. Section V shows how the method predicts operating parameter, cost, and/or performance goals for technologies without current sales potential (or for enhancing a frontier technology's sales potential). Section VI concludes with an illustrative example for the Chicago office building retrofit market

  1. High temperature friction and seizure in gas cooled nuclear reactors

    International Nuclear Information System (INIS)

    Cousseran, P.; Febvre, A.; Martin, R.; Roche, R.

    1978-01-01

    One of the most delicate problems encountered in the gas cooled nuclear reactors is the friction without lubrication in a dry and hot (800 0 C /1472 0 F) helium atmosphere even at very small velocity. The research and development programs are described together with special tribometers that operate at mode than 1000 0 C (1832 0 F) in dry helium. The most interesting test conditions and results are given for gas nitrited steels and for strongly alloyed Ni-Cr steels coated with chromium carbide by plasma sprayed. The effects of parameters live velocity, travelled distance, contact pressure, roughness, temperature and prolonged stops under charge are described together with the effects of negative phenomena like attachment and chattering [fr

  2. Gas cooled fast reactor background, facilities, industries and programmes

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1980-05-01

    This report was prepared at the request of the OECD-NEA Coordinating Group on Gas Cooled Fast Reactor Development and it represents a contribution (Vol.II) to the jointly sponsored Vol.I (GCFR Status Report). After a chapter on background with a brief description of the early studies and the activities in the various countries involved in the collaborative programme (Austria, Belgium, France, Germany, Japan, Sweden, Switzerland, United Kingdom and United States), the report describes the facilities available in those countries and at the Gas Breeder Reactor Association and the industrial capabilities relevant to the GCFR. Finally the programmes are described briefly with programme charts, conclusions and recommendations are given. (orig.) [de

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

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1979-06-01

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

  4. Multidisciplinary design optimization of film-cooled gas turbine blades

    Directory of Open Access Journals (Sweden)

    Talya Shashishekara S.

    1999-01-01

    Full Text Available Design optimization of a gas turbine blade geometry for effective film cooling toreduce the blade temperature has been done using a multiobjective optimization formulation. Three optimization formulations have been used. In the first, the average blade temperature is chosen as the objective function to be minimized. An upper bound constraint has been imposed on the maximum blade temperature. In the second, the maximum blade temperature is chosen as the objective function to be minimized with an upper bound constraint on the average blade temperature. In the third formulation, the blade average and maximum temperatures are chosen as objective functions. Shape optimization is performed using geometric parameters associated with film cooling and blade external shape. A quasi-three-dimensional Navier–Stokes solver for turbomachinery flows is used to solve for the flow field external to the blade with appropriate modifications to incorporate the effect of film cooling. The heat transfer analysis for temperature distribution within the blade is performed by solving the heat diffusion equation using the finite element method. The multiobjective Kreisselmeier–Steinhauser function approach has been used in conjunction with an approximate analysis technique for optimization. The results obtained using both formulations are compared with reference geometry. All three formulations yield significant reductions in blade temperature with the multiobjective formulation yielding largest reduction in blade temperature.

  5. Static and dynamic modelling of gas turbines in advanced cycles

    Energy Technology Data Exchange (ETDEWEB)

    Gustafsson, Jan-Olof

    1998-12-01

    Gas turbines have been in operation for at least 50 years. The engine is used for propulsion of aircraft and high speed ships. It is used for power production in remote locations and for peak load and emergency situations. Gas turbines have been used in combined cycles for 20 to 30 years. Highly efficient power plants based on gas turbines are a competitive option for the power industry today. The thermal efficiency of the simple cycle gas turbine has increased due to higher turbine inlet temperatures and improved compressor and expander designs. Equally important are the improved cycles in which the gas turbine operates. One example is the combined cycle that uses steam for turbine cooling. Steam is extracted from the bottoming cycle, then used as airfoil coolant in a closed loop and returned to the bottoming cycle. The Evaporative Gas Turbine (EvGT), also known as the Humid Air Turbine (HAT), is another advanced cycle. A mixture of air and water vapour is used as working media. Air from the compressor outlet is humidified and then preheated in a recuperator prior to combustion. The static and dynamic performance is changed when the gas turbine is introduced in an evaporative cycle. The cycle is gaining in popularity, but so far it has not been demonstrated. A Swedish joint program to develop the cycle has been in operation since 1993. As part of the program, a small pilot plant is being erected at the Lund Institute of Technology (LTH). The plant is based on a 600 kW gas turbine, and demonstration of the EvGT cycle started autumn 1998 and will continue, in the present phase, for one year. This thesis presents static and dynamic models for traditional gas turbine components, such as, the compressor, combustor, expander and recuperator. A static model for the humidifier is presented, based on common knowledge for atmospheric humidification. All models were developed for the pilot plant at LTH with the objective to support evaluation of the process and individual

  6. Advances in gas-liquid flows 1990

    International Nuclear Information System (INIS)

    Kim, J.M.; Hashemi, A.

    1990-01-01

    Gas-liquid two-phase flows commonly occur in nature and industrial applications. Rain, clouds, geysers, and waterfalls are examples of natural gas-liquid flow phenomena, whereas industrial applications can be found in nuclear reactors, steam generators, boilers, condensers, evaporators, fuel atomization, heat pipes, electronic equipment cooling, petroleum engineering, chemical process engineering, and many others. The household-variety phenomena such as garden sprinklers, shower, whirlpool bath, dripping faucet, boiling tea pot, and bubbling beer provide daily experience of gas-liquid flows. The papers presented in this volume reflect the variety and richness of gas-liquid two-phase flow and the increasing role it plays in modern technology. This volume contains papers dealing with some recent development in gas-liquid flow science and technology, covering basic gas-liquid flows, measurements and instrumentation, cavitation and flashing flows, countercurrent flow and flooding, flow in various components and geometries liquid metals and thermocapillary effects, heat transfer, nonlinear phenomena, instability, and other special and general topics related to gas-liquid flows

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

    Science.gov (United States)

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

    2013-09-03

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

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

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

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

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

    International Nuclear Information System (INIS)

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

    1976-10-01

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

  10. Modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1988-01-01

    The high financial risk involved in building large nuclear power reactors has been a major factor in halting investment in new plant and in bringing further technical development to a standstill. Increased public concern about the safety of nuclear plant, particularly after Chernobyl, has contributed to this stagnation. Financial and technical risk could be reduced considerably by going to small modular units, which would make it possible to build up power station capacity in small steps. Such modular plant, based on the helium-cooled high temperature reactor (HTR), offers remarkable advantages in terms of inherent safety characteristics, partly because of the relatively small size of the individual modules but more on account of the enormous thermal capacity and high temperature margins of the graphitic reactor assemblies. Assessments indicate that, in the USA, the cost of power from the modular systems would be less than that from conventional single reactor plant, up to about 600 MW(e), and only marginally greater above that level, a margin that should be offset by the shorter time required in bringing the modular units on line to earn revenue. The modular HTR would be particularly appropriate in the UK, because of the considerable British industrial background in gas-cooled reactors, and could be a suitable replacement for Magnox. The modular reactor would be particularly suited to combined heat and power schemes and would offer great potential for the eventual development of gas turbine power conversion and the production of high-temperature process heat. (author)

  11. Evaluation of the RSG-GAS cooling tower performance

    International Nuclear Information System (INIS)

    Suroso

    2003-01-01

    Utilization of RSG-GAS reactor should be operated as efficiently as possible, so that reactor operation planning using one line primary coolant can be anticipated. To analyze the performance of the RSG-GAS cooling tower with one line primary coolant doing by using same data from 10 MW thermal reactor operation. The result were then compare to those achieved using CATHENA code. The results indicated that, for design condition the ratio of water flowrate to air is (L/G) 1.52 and number transfer unit (NTU) is 0.348. For operation condition, the average of L/G and NTU are respectively 1.37 and 0,348. Moreover the results achieved by the code showed that L/G and NTU are respectively 1.35 and 0,302. The performance of cooling tower achieved operation condition and the code results are respectively 91% and 72%. This means that the calculated results are lower than measurement results

  12. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

  13. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  14. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  15. A review of helium gas turbine technology for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    No, Hee Cheon; Kim, Ji Hwan; Kim, Hyeun Min

    2007-01-01

    Current High-Temperature Gas-cooled Reactors (HTGRs) are based on a closed brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference

  16. Combustion modeling in advanced gas turbine systems

    Energy Technology Data Exchange (ETDEWEB)

    Smoot, L.D.; Hedman, P.O.; Fletcher, T.H. [Brigham Young Univ., Provo, UT (United States)] [and others

    1995-10-01

    The goal of the U.S. Department of Energy`s Advanced Turbine Systems (ATS) program is to help develop and commercialize ultra-high efficiency, environmentally superior, and cost competitive gas turbine systems for base-load applications in the utility, independent power producer, and industrial markets. Combustion modeling, including emission characteristics, has been identified as a needed, high-priority technology by key professionals in the gas turbine industry.

  17. Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lee, Won-Jae

    2006-01-01

    The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

  18. Preliminary analysis of combined cycle of modular high-temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Baogang, Z.; Xiaoyong, Y.; Jie, W.; Gang, Z.; Qian, S.

    2015-01-01

    Modular high-temperature gas cooled reactor (HTGR) is known as one of the most advanced nuclear reactors because of its inherent safety and high efficiency. The power conversion system of HTGR can be steam turbine based on Rankine cycle or gas turbine based on Brayton cycle respectively. The steam turbine system is mature and the gas turbine system has high efficiency but under development. The Brayton-Rankine combined cycle is an effective way to further promote the efficiency. This paper investigated the performance of combined cycle from the viewpoint of thermodynamics. The effect of non-dimensional parameters on combined cycle’s efficiency, such as temperature ratio, compression ratio, efficiency of compressor, efficiency of turbine, was analyzed. Furthermore, the optimal parameters to achieve highest efficiency was also given by this analysis under engineering constraints. The conclusions could be helpful to the design and development of combined cycle of HTGR. (author)

  19. High temperature gas cooled reactor technology development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-12-01

    The successful introduction of an advanced nuclear power plant programme depends on many key elements. It must be economically competitive with alternative sources of energy, its technical development must assure operational dependability, the support of society requires that it be safe and environmentally acceptable, and it must meet the regulatory standards developed for its use and application. These factors interrelate with each other, and the ability to satisfy the established goals and criteria of all of these requirements is mandatory if a country or a specific industry is to proceed with a new, advanced nuclear power system. It was with the focus on commercializing the high temperature gas cooled reactor (HTGR) that the IAEA's International Working Group on Gas Cooled Reactors recommended this Technical Committee Meeting (TCM) on HTGR Technology Development. Over the past few years, many Member States have instituted a re-examination of their nuclear power policies and programmes. It has become evident that the only realistic way to introduce an advanced nuclear power programme in today's world is through international co-operation between countries. The sharing of expertise and technical facilities for the common development of the HTGR is the goal of the Member States comprising the IAEA's International Working Group on Gas Cooled Reactors. This meeting brought together key representatives and experts on the HTGR from the national organizations and industries of ten countries and the European Commission. The state electric utility of South Africa, Eskom, hosted this TCM in Johannesburg, from 13 to 15 November 1996. This TCM provided the opportunity to review the status of HTGR design and development activities, and especially to identify international co-operation which could be utilized to bring about the commercialization of the HTGR

  20. Advanced Natural Gas Reciprocating Engine(s)

    Energy Technology Data Exchange (ETDEWEB)

    Kwok, Doris; Boucher, Cheryl

    2009-09-30

    Energy independence and fuel savings are hallmarks of the nation’s energy strategy. The advancement of natural gas reciprocating engine power generation technology is critical to the nation’s future. A new engine platform that meets the efficiency, emissions, fuel flexibility, cost and reliability/maintainability targets will enable American manufacturers to have highly competitive products that provide substantial environmental and economic benefits in the US and in international markets. Along with Cummins and Waukesha, Caterpillar participated in a multiyear cooperative agreement with the Department of Energy to create a 50% efficiency natural gas powered reciprocating engine system with a 95% reduction in NOx emissions by the year 2013. This platform developed under this agreement will be a significant contributor to the US energy strategy and will enable gas engine technology to remain a highly competitive choice, meeting customer cost of electricity targets, and regulatory environmental standard. Engine development under the Advanced Reciprocating Engine System (ARES) program was divided into phases, with the ultimate goal being approached in a series of incremental steps. This incremental approach would promote the commercialization of ARES technologies as soon as they emerged from development and would provide a technical and commercial foundation of later-developing technologies. Demonstrations of the Phase I and Phase II technology were completed in 2004 and 2008, respectively. Program tasks in Phase III included component and system development and testing from 2009-2012. Two advanced ignition technology evaluations were investigated under the ARES program: laser ignition and distributed ignition (DIGN). In collaboration with Colorado State University (CSU), a laser ignition system was developed to provide ignition at lean burn and high boost conditions. Much work has been performed in Caterpillar’s DIGN program under the ARES program. This work

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

    International Nuclear Information System (INIS)

    Winkleblack, R.K.

    1976-04-01

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

  2. High-temperature gas-cooled reactor (HTGR): long term program plan

    International Nuclear Information System (INIS)

    1980-01-01

    The FY 1980 effort was to investigate four technology options identified by program participants as potentially viable candidates for near-term demonstration: the Gas Turbine system (HTGR-GT), reflecting its perceived compatibility with the dry-cooling market, two systems addressing the process heat market, the Reforming (HTGR-R) and Steam Cycle (HTGR-SC) systems, and a more developmental reactor system, The Nuclear Heat Source Demonstration Reactor (NHSDR), which was to serve as a basis for both the HTGR-GT and HTGR-R systems as well as the further potential for developing advanced applications such as steam-coal gasification and water splitting

  3. Water cooled metal optics for the Advanced Light Source

    International Nuclear Information System (INIS)

    McKinney, W.R.; Irick, S.C.; Lunt, D.L.J.

    1991-01-01

    The program for providing water cooled metal optics for the Advanced Light Source at Berkeley is reviewed with respect to fabrication and metrology of the surfaces. Materials choices, surface figure and smoothness specifications, and metrology systems for measuring the plated metal surfaces are discussed. Results from prototype mirrors and grating blanks will be presented, which show exceptionally low microroughness and mid-period error. We will briefly describe out improved version of the Long Trace Profiler, and its importance to out metrology program. We have completely redesigned the mechanical, optical and computational parts of the profiler system with the cooperation of Peter Takacs of Brookhaven, Continental Optical, and Baker Manufacturing. Most important is that one of our profilers is in use at the vendor to allow testing during fabrication. Metrology from the first water cooled mirror for an ALS beamline is presented as an example. The preplating processing and grinding and polishing were done by Tucson Optical. We will show significantly better surface microroughness on electroless nickel, over large areas, than has been reported previously

  4. Advanced Spectral Library (ASTRAL): Atomic Fluorescence in Cool, Evolved Stars

    Science.gov (United States)

    Carpenter, Ken G.; Nielsen, Krister E.; Kober, Gladys V.; Rau, Gioia

    2018-01-01

    The "Advanced Spectral Library (ASTRAL) Project: Cool Stars" (PI = T. Ayres) collected a definitive set of representative, high-resolution (R~46,000 in the FUV up to ~1700 Å, R~30,000 for 1700-2150 Å, and R~114,000 >2150 Å) and high signal/noise (S/N>100) UV spectra of eight F-M evolved cool stars. These extremely high-quality STIS UV echelle spectra are available from the HST archive and from the Univ. of Colorado (http://casa.colorado.edu/~ayres/ASTRAL/) and will enable investigations of a broad range of problems -- stellar, interstellar, and beyond -- for many years. In this paper, we extend our study of the very rich emission-line spectra of the four evolved K-M stars in the sample, Beta Gem (K0 IIIb), Gamma Dra (K5 III), Gamma Cru (M3.4 III), and Alpha Ori (M2 Iab), to study the atomic fluorescence processes operating in their outer atmospheres. We summarize the pumping transitions and fluorescent line products known on the basis of previous work (e.g. Carpenter 1988, etc.) and newly identified in our current, on-going analysis of these extraordinary ASTRAL STIS spectra.

  5. Challenges in licensing a sodium-cooled advanced recycling reactor

    International Nuclear Information System (INIS)

    Levin, Alan E.

    2008-01-01

    As part of the Global Nuclear Energy Partnership (GNEP), the U.S. Department of Energy (DOE) has focused on the use of sodium-cooled fast reactors (SFRs) for the destruction of minor actinides derived from used reactor fuel. This approach engenders an array of challenges with respect to the licensing of the reactor: the U.S. Nuclear Regulatory Commission (NRC) has never completed the review of an application for an operating license for a sodium-cooled reactor. Moreover, the current U.S. regulatory structure has been developed to deal almost exclusively with light-water reactor (LWR) designs. Consequently, the NRC must either (1) develop a new regulatory process for SFRs, or (2) reinterpret the existing regulations to apply them, as appropriate, to SFR designs. During the 1980s and 1990s, the NRC conducted preliminary safety assessments of the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Innovative Small Module (PRISM) designs, and in that context, began to consider how to apply LWR-based regulations to SFR designs. This paper builds on that work to consider the challenges, from the reactor designer's point of view, associated with licensing an SFR today, considering (1) the evolution of SFR designs, (2) the particular requirements of reactor designs to meet GNEP objectives, and (3) the evolution of NRC regulations since the conclusion of the SAFR and PRISM reviews. (author)

  6. Cryogenically cooled monochromators for the Advanced Photon Source

    International Nuclear Information System (INIS)

    Mills, D.M.

    1996-01-01

    The use of cryogenically cooled monochromators looks to be a very promising possibility for the Advanced Photon Source. This position has recently been bolstered by several experiments performed on beamlines at the ESRF and CHESS. At the ESRF, several crystal geometries have been tested that were designed for high power densities (approx-gt 150 W/mm 2 ) and moderate total absorbed powers (<200 W). These geometries have proven to be very successful at handling these power parameters with measured strains on the arc-second level. The experiments performed at CHESS were focused on high total power (approx-gt 1000 W) but moderate power densities. As with the previously mentioned experiments, the crystals designed for this application performed superbly with no measurable broadening of the rocking curves on the arc-second level. These experiments will be summarized and, based on these results, the performance of cryogenic monochromators for the APS will be assessed. copyright 1996 American Institute of Physics

  7. Auxiliary bearing design considerations for gas cooled reactors

    International Nuclear Information System (INIS)

    Penfield, S.R. Jr.; Rodwell, E.

    2001-01-01

    The need to avoid contamination of the primary system, along with other perceived advantages, has led to the selection of electromagnetic bearings (EMBs) in most ongoing commercial-scale gas cooled reactor (GCR) designs. However, one implication of magnetic bearings is the requirement to provide backup support to mitigate the effects of failures or overload conditions. The demands on these auxiliary or 'catcher' bearings have been substantially escalated by the recent development of direct Brayton cycle GCR concepts. Conversely, there has been only limited directed research in the area of auxiliary bearings, particularly for vertically oriented turbomachines. This paper explores the current state-of-the-art for auxiliary bearings and the implications for current GCR designs. (author)

  8. Development of high temperature gas cooled reactor in China

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

    2018-02-15

    High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

  9. Acoustical environment of gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Blevins, R.D.

    1986-01-01

    Methods for acoustical analysis of gas-cooled nuclear reactors in terms of the sources of sound, the propagation of sound about the coolant circuit and the response of reactor structures to sound, are described. Sources of sound that are considered are circulators, jets, vortex shedding and separated flow. Circulators are generally the dominant source of sound. At low frequency the sound propagates one dimensionally through the ducts and cavities of the reactor. At high frequency the sound excites closely spaced two- and three-dimensional acoustic modes, and the resultant sound field can be described only statistically. The sound excites plate and shell structures within the coolant circuit. Secondary steam piping can also be excited by pumps and valves. Formulations are presented for the resultant vibration. Vibration-induced damage is also reviewed. (author)

  10. Safety analysis of a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimazu, Akira; Morimoto, Toshio

    1975-01-01

    In recent years, in order to satisfy the social requirements of environment and safety and also to cope with the current energy stringency, the installation of safe nuclear power plants is indispensable. Herein, safety analysis and evaluation to confirm quantitatively the safety design of a nuclear power plant become more and more important. The safety analysis and its methods for a high temperature gas-cooled reactor are described, with emphasis placed on the practices by Fuji Electric Manufacturing Co. Fundamental rule of securing plant safety ; safety analysis in normal operation regarding plant dynamic characteristics and radioactivity evaluation ; and safety analysis at the time of accidents regarding plant response to the accidents and radioactivity evaluation are explained. (Mori, K.)

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

    International Nuclear Information System (INIS)

    Jensen, S.E.; Nonboel, E.

    1999-05-01

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

  12. A He-gas Cooled, Stationary Granular Target

    CERN Document Server

    Pugnat, P

    2003-01-01

    In the CERN approach to the design of a neutrino factory, the repetition frequency of the proton beam is high enough to consider stationary solid targets as a viable solution for multi-MW beams. The target consists of high density tantalum spheres of 2 mm diameter which can efficiently be cooled by passing a high mass flow He-gas stream through the voids between the Ta-granules. Very small thermal shocks and stresses will arise in this fine grained structure due to the relatively long burst of 3.3 ms from the SPL-proton linac. In a quadruple target system where each target receives only one quarter of the total beam power of 4 MW, conservative temperature levels and adequate lifetimes of the target are estimated in its very high radiation environment. A conceptual design of the integration of the target into the magnetic horn-pion-collector is presented.

  13. Development of failure detection system for gas-cooled reactor

    International Nuclear Information System (INIS)

    Feirreira, M.P.

    1990-01-01

    This work presents several kinds of Failure Detection Systems for Fuel Elements, stressing their functional principles and major applications. A comparative study indicates that the method of electrostatic precipitation of the fission gases Kr and Xe is the most efficient for fuel failure detection in gas-cooled reactors. A detailed study of the physical phenomena involved in electrostatic precipitation led to the derivation of an equation for the measured counting rate. The emission of fission products from the fuel and the ion recombination inside the chamber are evaluated. A computer program, developed to simulate the complete operation of the system, relates the counting rate to the concentration of Kr and Xe isotopes. The project of a mock-up is then presented. Finally, the program calculations are compared to experimental data, available from the literature, yielding a close agreement. (author)

  14. Gas-cooled reactor commercialization study. Interim report

    International Nuclear Information System (INIS)

    1977-01-01

    This report of the gas-cooled reactor commercialization study completes the technical and cost evaluation portions of this study contract. A final report in December will update the status of the incentive analyses and the issues of commercialization. This study was designed to bring together potential industry participants (utilities and suppliers) to evaluate the commercial potential of the HTGR-SC and to build channels of communication among the participating organizations at the same time that technical, economic and institutional issues were being evaluated. RAMCO, Inc., in suggesting and using this study approach, believes its application extends to any commercialization problem involving multi-party involvement in high capital, intensive, high risk energy technologies

  15. Study on plant concept for gas cooled fast reactor

    International Nuclear Information System (INIS)

    Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

    2003-05-01

    In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

  16. Metaphysics methods development for high temperature gas cooled reactor analysis

    International Nuclear Information System (INIS)

    Seker, V.; Downar, T. J.

    2007-01-01

    Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology road map. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark

  17. Material System Engineering for Advanced Electrocaloric Cooling Technology

    Science.gov (United States)

    Qian, Xiaoshi

    Electrocaloric effect refers to the entropy change and/or temperature change in dielectrics caused by the electric field induced polarization change. Recent discovery of giant ECE provides an opportunity to realize highly efficient cooling devices for a broad range of applications ranging from household appliances to industrial applications, from large-scale building thermal management to micro-scale cooling devices. The advances of electrocaloric (EC) based cooling device prototypes suggest that highly efficient cooling devices with compact size are achievable, which could lead to revolution in next generation refrigeration technology. This dissertation focuses on both EC based materials and cooling devices with their recent advances that address practical issues. Based on better understandings in designing an EC device, several EC material systems are studied and improved to promote the performances of EC based cooling devices. In principle, applying an electric field to a dielectric would cause change of dipolar ordering states and thus a change of dipolar entropy. Giant ECE observed in ferroelectrics near ferroelectric-paraelectric (FE-PE) transition temperature is owing to the large dipolar orientation change, between random-oriented dipolar states in paraelectric phase and spontaneous-ordered dipolar states in ferroelectric phases, which is induced by external electric fields. Besides pursuing large ECE, studies on EC cooling devices indicated that EC materials are required to possess wide operational temperature window, in which large ECE can be maintained for efficient operations. Although giant ECE was first predicted in ferroelectric polymers, where the large effect exhibits near FEPE phase transition, the narrow operation temperature window poses obstacles for these normal ferroelectrics to be conveniently perform in wide range of applications. In this dissertation, we demonstrated that the normal ferroelectric polymers can be converted to relaxor

  18. Advanced Pumps and Cold Plates for Two-Phase Cooling Loops, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Advanced instruments used for earth science missions require improved cooling systems to remove heat from high power electronic components and maintain tight...

  19. Assessment of gas cooled fast reactor with indirect supercritical CO2 cycle

    International Nuclear Information System (INIS)

    Hejzlar, P.; Driscoll, M. J.; Dostal, V.; Dumaz, P.; Poullennec, G.; Alpy, N.

    2006-01-01

    Various indirect power cycle options for a helium cooled Gas cooled Fast Reactor (GFR) with particular focus on a supercritical CO 2 (SCO 2 ) indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The Balance Of Plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and SCO 2 recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of 550 .deg. C, (2) advanced design with turbine inlet temperature of 650 .deg. C and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect SCO 2 recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR 'proximate-containment' and the BOP for the SCO 2 cycle is very compact. Both these factors will lead to reduced capital cost

  20. Assessment and status report High-Temperature Gas-Cooled Reactor gas-turbine technology

    International Nuclear Information System (INIS)

    1981-01-01

    Purpose of this report is to present a brief summary assessment of the High Temperature Gas-Cooled Reactor - Gas Turbine (HTGR-GT) technology. The focal point for the study was a potential 2000 MW(t)/800 MW(e) HTGR-GT commercial plant. Principal findings of the study were that: the HTGR-GT is feasible, but with significantly greater development risk than the HTGR-SC (Steam Cycle). At the level of performance corresponding to the reference design, no incremental economic incentive can be identified for the HTGR-GT to offset the increased development costs and risk relative to the HTGR-SC. The relative economics of the HTGR-GT and HTGR-SC are not significantly impacted by dry cooling considerations. While reduced cycel complexity may ultimately result in a reliability advantage for the HTGR-GT, the value of that potential advantage was not quantified

  1. Low pressure cooling seal system for a gas turbine engine

    Science.gov (United States)

    Marra, John J

    2014-04-01

    A low pressure cooling system for a turbine engine for directing cooling fluids at low pressure, such as at ambient pressure, through at least one cooling fluid supply channel and into a cooling fluid mixing chamber positioned immediately downstream from a row of turbine blades extending radially outward from a rotor assembly to prevent ingestion of hot gases into internal aspects of the rotor assembly. The low pressure cooling system may also include at least one bleed channel that may extend through the rotor assembly and exhaust cooling fluids into the cooling fluid mixing chamber to seal a gap between rotational turbine blades and a downstream, stationary turbine component. Use of ambient pressure cooling fluids by the low pressure cooling system results in tremendous efficiencies by eliminating the need for pressurized cooling fluids for sealing this gap.

  2. Thermodynamic assessment of impact of inlet air cooling techniques on gas turbine and combined cycle performance

    International Nuclear Information System (INIS)

    Mohapatra, Alok Ku; Sanjay

    2014-01-01

    The article is focused on the comparison of impact of two different methods of inlet air cooling (vapor compression and vapor absorption cooling) integrated to a cooled gas turbine based combined cycle plant. Air-film cooling has been adopted as the cooling technique for gas turbine blades. A parametric study of the effect of compressor pressure ratio, compressor inlet temperature (T i , C ), turbine inlet temperature (T i , T ), ambient relative humidity and ambient temperature on performance parameters of plant has been carried out. Optimum T i , T corresponding to maximum plant efficiency of combined cycle increases by 100 °C due to the integration of inlet air cooling. It has been observed that vapor compression cooling improves the efficiency of gas turbine cycle by 4.88% and work output by 14.77%. In case of vapor absorption cooling an improvement of 17.2% in gas cycle work output and 9.47% in gas cycle efficiency has been observed. For combined cycle configuration, however, vapor compression cooling should be preferred over absorption cooling in terms of higher plant performance. The optimum value of compressor inlet temperature has been observed to be 20 °C for the chosen set of conditions for both the inlet air cooling schemes. - Highlights: • Inlet air cooling improves performance of cooled gas turbine based combined cycle. • Vapor compression inlet air cooling is superior to vapor absorption inlet cooling. • For every turbine inlet temperature, there exists an optimum pressure ratio. • The optimum compressor inlet temperature is found to be 293 K

  3. Present state and future prospect of development of high temperature gas-cooled reactors in Japan

    International Nuclear Information System (INIS)

    Sanokawa, Konomo

    1994-01-01

    High temperature gas-cooled reactors can supply the heat of about 1000degC, and the high efficiency and the high rate of heat utilization can be attained. Also they have the features of excellent inherent safety, the easiness of operation, the high burnup of fuel and so on. The heat utilization of atomic energy in addition to electric power generation is very important in view of the protection of global environment and the diversification of energy supply. Japan Atomic Energy Research Institute has advanced the construction of the high temperature engineering test and research reactor (HTTR) of 30 MW thermal output, aiming at attaining the criticality in 1998. The progress of the development of a high temperature gas-cooled reactor is described. For 18 years, the design study of the reactor was advanced together with the research and development of the reactor physics, fuel and materials, high temperature machinery and equipment and others, and the decision of the design standard and the development of computation codes. The main specification and the construction schedule are shown. The reactor building was almost completed, and the reactor containment vessel was installed. The plan of the research and development by using the HTTR is investigated. (K.I.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  5. Gas expulsion vs gas retention in young stellar clusters II: effects of cooling and mass segregation

    Science.gov (United States)

    Silich, Sergiy; Tenorio-Tagle, Guillermo

    2018-05-01

    Gas expulsion or gas retention is a central issue in most of the models for multiple stellar populations and light element anti-correlations in globular clusters. The success of the residual matter expulsion or its retention within young stellar clusters has also a fundamental importance in order to understand how star formation proceeds in present-day and ancient star-forming galaxies and if proto-globular clusters with multiple stellar populations are formed in the present epoch. It is usually suggested that either the residual gas is rapidly ejected from star-forming clouds by stellar winds and supernova explosions, or that the enrichment of the residual gas and the formation of the second stellar generation occur so rapidly, that the negative stellar feedback is not significant. Here we continue our study of the early development of star clusters in the extreme environments and discuss the restrictions that strong radiative cooling and stellar mass segregation provide on the gas expulsion from dense star-forming clouds. A large range of physical initial conditions in star-forming clouds which include the star-forming cloud mass, compactness, gas metallicity, star formation efficiency and effects of massive stars segregation are discussed. It is shown that in sufficiently massive and compact clusters hot shocked winds around individual massive stars may cool before merging with their neighbors. This dramatically reduces the negative stellar feedback, prevents the development of the global star cluster wind and expulsion of the residual and the processed matter into the ambient interstellar medium. The critical lines which separate the gas expulsion and the gas retention regimes are obtained.

  6. Validation of CATHARE for gas-cooled reactors

    International Nuclear Information System (INIS)

    Fabrice Bentivoglio; Ola Widlund; Manuel Saez

    2005-01-01

    Full text of publication follows: Extensively validated and qualified for light-water reactor safety studies, the thermo-hydraulics code CATHARE has been adapted to deal also with gas-cooled reactor applications. In order to validate the code for these novel applications, CEA (Commissariat a l'Energie Atomique) has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE is being validated against existing experimental data, in particular from the German power plant Oberhausen II and the South African Pebble-Bed Micro Model (PBMM). Oberhausen II, operated by the German utility EVO, is a 50 MW(e) direct-cycle Helium turbine plant. The power source is a gas burner rather than a nuclear reactor core, but the power conversion system resembles those of the GFR (Gas-cooled Fast Reactor) and other high-temperature reactor concepts. Oberhausen II was operated for more than 100 000 hours between 1974 and 1988. Design specifications, drawings and experimental data have been obtained through the European HTR project, offering a unique opportunity to validate CATHARE on a large-scale Brayton cycle. Available measurements of temperatures, pressures and mass flows throughout the circuit have allowed a very comprehensive thermohydraulic description of the plant, in steady-state conditions as well as during transients. The Pebble-Bed Micro Model (PBMM) is a small-scale model conceived to demonstrate the operability and control strategies of the South African PBMR concept. The model uses Nitrogen instead of Helium, and an electrical heater with a maximum rating of 420 kW. As the full-scale PBMR, the PBMM loop features three turbines and two compressors on the primary circuit, located on three separate shafts. The generator, however, is modelled by a third compressor on a separate circuit, with a

  7. Nuclear closed-cycle gas turbine (HTGR-GT): dry cooled commercial power plant studies

    International Nuclear Information System (INIS)

    McDonald, C.F.; Boland, C.R.

    1979-11-01

    Combining the modern and proven power conversion system of the closed-cycle gas turbine (CCGT) with an advanced high-temperature gas-cooled reactor (HTGR) results in a power plant well suited to projected utility needs into the 21st century. The gas turbine HTGR (HTGR-GT) power plant benefits are consistent with national energy goals, and the high power conversion efficiency potential satisfies increasingly important resource conservation demands. Established technology bases for the HTGR-GT are outlined, together with the extensive design and development program necessary to commercialize the nuclear CCGT plant for utility service in the 1990s. This paper outlines the most recent design studies by General Atomic for a dry-cooled commercial plant of 800 to 1200 MW(e) power, based on both non-intercooled and intercooled cycles, and discusses various primary system aspects. Details are given of the reactor turbine system (RTS) and on integrating the major power conversion components in the prestressed concrete reactor vessel

  8. The influence of liquid-gas velocity ratio on the noise of the cooling tower

    Science.gov (United States)

    Yang, Bin; Liu, Xuanzuo; Chen, Chi; Zhao, Zhouli; Song, Jinchun

    2018-05-01

    The noise from the cooling tower has a great influence on psychological performance of human beings. The cooling tower noise mainly consists of fan noise, falling water noise and mechanical noise. This thesis used DES turbulence model with FH-W model to simulate the flow and sound pressure field in cooling tower based on CFD software FLUENT and analyzed the influence of different kinds noise, which affected by diverse factors, on the cooling tower noise. It can be concluded that the addition of cooling water can reduce the turbulence and vortex noise of the rotor fluid field in the cooling tower at some extent, but increase the impact noise of the liquid-gas two phase. In general, the cooling tower noise decreases with the velocity ratio of liquid to gas increasing, and reaches the lowest when the velocity ratio of liquid to gas is close to l.

  9. Advanced multistage turbine blade aerodynamics, performance, cooling, and heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Fleeter, S.; Lawless, P.B. [Purdue Univ., West Lafayette, IN (United States)

    1995-10-01

    The gas turbine has the potential for power production at the highest possible efficiency. The challenge is to ensure that gas turbines operate at the optimum efficiency so as to use the least fuel and produce minimum emissions. A key component to meeting this challenge is the turbine. Turbine performance, both aerodynamics and heat transfer, is one of the barrier advanced gas turbine development technologies. This is a result of the complex, highly three-dimensional and unsteady flow phenomena in the turbine. Improved turbine aerodynamic performance has been achieved with three-dimensional highly-loaded airfoil designs, accomplished utilizing Euler or Navier-Stokes Computational Fluid Dynamics (CFD) codes. These design codes consider steady flow through isolated blade rows. Thus they do not account for unsteady flow effects. However, unsteady flow effects have a significant impact on performance. Also, CFD codes predict the complete flow field. The experimental verification of these codes has traditionally been accomplished with point data - not corresponding plane field measurements. Thus, although advanced CFD predictions of the highly complex and three-dimensional turbine flow fields are available, corresponding data are not. To improve the design capability for high temperature turbines, a detailed understanding of the highly unsteady and three-dimensional flow through multi-stage turbines is necessary. Thus, unique data are required which quantify the unsteady three-dimensional flow through multi-stage turbine blade rows, including the effect of the film coolant flow. This requires experiments in appropriate research facilities in which complete flow field data, not only point measurements, are obtained and analyzed. Also, as design CFD codes do not account for unsteady flow effects, the next logical challenge and the current thrust in CFD code development is multiple-stage analyses that account for the interactions between neighboring blade rows.

  10. The early history of high-temperature helium gas-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Simnad, M.T.; California Univ., San Diego, La Jolla, CA

    1991-01-01

    The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

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

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, P.J.

    1997-04-01

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

  13. Advanced coal-fueled gas turbine systems

    Energy Technology Data Exchange (ETDEWEB)

    Wenglarz, R.A.

    1994-08-01

    Several technology advances since the early coal-fueled turbine programs that address technical issues of coal as a turbine fuel have been developed in the early 1980s: Coal-water suspensions as fuel form, improved methods for removing ash and contaminants from coal, staged combustion for reducing NO{sub x} emissions from fuel-bound nitrogen, and greater understanding of deposition/erosion/corrosion and their control. Several Advanced Coal-Fueled Gas Turbine Systems programs were awarded to gas turbine manufacturers for for components development and proof of concept tests; one of these was Allison. Tests were conducted in a subscale coal combustion facility and a full-scale facility operating a coal combustor sized to the Allison Model 501-K industrial turbine. A rich-quench-lean (RQL), low nitrogen oxide combustor design incorporating hot gas cleanup was developed for coal fuels; this should also be applicable to biomass, etc. The combustor tests showed NO{sub x} and CO emissions {le} levels for turbines operating with natural gas. Water washing of vanes from the turbine removed the deposits. Systems and economic evaluations identified two possible applications for RQL turbines: Cogeneration plants based on Allison 501-K turbine (output 3.7 MW(e), 23,000 lbs/hr steam) and combined cycle power plants based on 50 MW or larger gas turbines. Coal-fueled cogeneration plant configurations were defined and evaluated for site specific factors. A coal-fueled turbine combined cycle plant design was identified which is simple, compact, and results in lower capital cost, with comparable efficiency and low emissions relative to other coal technologies (gasification, advanced PFBC).

  14. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  15. The first high resolution image of coronal gas in a starbursting cool core cluster

    Science.gov (United States)

    Johnson, Sean

    2017-08-01

    Galaxy clusters represent a unique laboratory for directly observing gas cooling and feedback due to their high masses and correspondingly high gas densities and temperatures. Cooling of X-ray gas observed in 1/3 of clusters, known as cool-core clusters, should fuel star formation at prodigious rates, but such high levels of star formation are rarely observed. Feedback from active galactic nuclei (AGN) is a leading explanation for the lack of star formation in most cool clusters, and AGN power is sufficient to offset gas cooling on average. Nevertheless, some cool core clusters exhibit massive starbursts indicating that our understanding of cooling and feedback is incomplete. Observations of 10^5 K coronal gas in cool core clusters through OVI emission offers a sensitive means of testing our understanding of cooling and feedback because OVI emission is a dominant coolant and sensitive tracer of shocked gas. Recently, Hayes et al. 2016 demonstrated that synthetic narrow-band imaging of OVI emission is possible through subtraction of long-pass filters with the ACS+SBC for targets at z=0.23-0.29. Here, we propose to use this exciting new technique to directly image coronal OVI emitting gas at high resolution in Abell 1835, a prototypical starbursting cool-core cluster at z=0.252. Abell 1835 hosts a strong cooling core, massive starburst, radio AGN, and at z=0.252, it offers a unique opportunity to directly image OVI at hi-res in the UV with ACS+SBC. With just 15 orbits of ACS+SBC imaging, the proposed observations will complete the existing rich multi-wavelength dataset available for Abell 1835 to provide new insights into cooling and feedback in clusters.

  16. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

  17. Use of thorium for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Guimarães, Cláudio Q., E-mail: claudio_guimaraes@usp.br [Universidade de São Paulo (USP), SP (Brazil). Instituto de Física; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The HTGR ( High Temperature Gas-cooled Reactor) is a 4{sup th} generation nuclear reactor and is fuelled by a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: The German “pebble bed” mode, which is a system that uses spherical fuel elements, containing a graphite-and-fuel mixture coated in a graphite shell; and the American version, whose fuel is loaded into precisely located graphite hexagonal prisms that interlock to create the core of the vessel. In both variants, the coolant consists of helium pressurised. The HTGR system operates most efficiently with the thorium fuel cycle, however, so relatively little development has been carried out in this country on that cycle for HTGRs. In the Nuclear Engineering Centre of IPEN (Instituto de Pesquisas Energéticas e Nucleares), a study group is being formed linked to thorium reactors, whose proposal is to investigate reactors using thorium for {sup 233}U production and rejects burning. The present work intends to show the use of thorium in HTGRs, their advantages and disadvantages and its feasibility. (author)

  18. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  19. Modelling the radiolysis of RSG-GAS primary cooling water

    Science.gov (United States)

    Butarbutar, S. L.; Kusumastuti, R.; Subekti, M.; Sunaryo, G. R.

    2018-02-01

    Water chemistry control for light water coolant reactor required a reliable understanding of radiolysis effect in mitigating corrosion and degradation of reactor structure material. It is known that oxidator products can promote the corrosion, cracking and hydrogen pickup both in the core and in the associated piping components of the reactor. The objective of this work is to provide the radiolysis model of RSG GAS cooling water and further more to predict the oxidator concentration which can lead to corrosion of reactor material. Direct observations or measurements of the chemistry in and around the high-flux core region of a nuclear reactor are difficult due to the extreme conditions of high temperature, pressure, and mixed radiation fields. For this reason, chemical models and computer simulations of the radiolysis of water under these conditions are an important route of investigation. FACSIMILE were used to calculate the concentration of O2 formed at relatively long-time by the pure water γ and neutron irradiation (pH=7) at temperature between 25 and 50 °C. This simulation method is based on a complex chemical reaction kinetic. In this present work, 300 MeV-proton were used to mimic γ-rays radiolysis and 2 MeV fast neutrons. Concentration of O2 were calculated at 10-6 - 106 s time scale.

  20. New deployment of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Tsuchie, Yasuo; Kunitomi, Kazuhiko; Shiozawa, Shusaku; Konuki, Kaoru; Inagaki, Yoshiyuki; Hayakawa, Hitoshi

    2002-01-01

    The high temperature gas-cooled reactor (HTGR) is now under a condition difficult to know it well, because of considering not only power generation, but also diverse applications of its nuclear heat, of having extremely different safe principle from that of conventional reactors, of having two types of pebble-bed and block which are extremely different types, of promoting its construction plan in South Africa, of including its application to disposition of Russian surplus weapons plutonium of less reporting HTTR in Japan in spite of its full operation, and so on. However, HTGR is expected for an extremely important nuclear reactor aiming at the next coming one of LWR. HTGR which is late started and developed under complete private leading, is strongly conscious at environmental problem since its beginning. Before 30 years when large scale HTGR was expected to operate, it advertised a merit to reduce wasted heat because of its high temperature. As ratio occupied by electricity expands among application of energies, ratio occupied by the other energies are larger. When considering applications except electric power, high temperature thermal energy from HTGR can be thought wider applications than that from LWR and so on. (G.K.)

  1. Inherently safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yamada, Masao; Hayakawa, Hitoshi

    1987-01-01

    It is recognized in general that High Temperature Gas-cooled Reactors have remarkable characteristics in inherent safety and it is well known that credits of the time margin have been admitted for accident evaluation in the licensing of the currently operating prototype HTGRs (300 MWe class). Recently, more inherently safe HTGRs are being developed in various countries and drawing attention on their possibility for urban siting. The inherent safety characteristics of these HTRs differ each other depending on their design philosophy and on the features of the components/structures which constitute the plant. At first, the specific features/characteristics of the elemental components/structures of the HTRs are explained one by one and then the overall safety features/characteristics of these HTR plants are explained in connection with their design philosophy and combination of the elemental features. Taking the KWU/Interatom Modular Reactor System as an example, the particular design philosophy and safety characteristics of the inherently safe HTR are explained with a result of preliminary evaluation on the possibility of siting close to densely populated area. (author)

  2. Unsteady thermal analysis of gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    Lakkis, I.A.

    1993-01-01

    This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously

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

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

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

  4. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    International Nuclear Information System (INIS)

    Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

    2004-01-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  5. Performance comparison of liquid metal and gas cooled ATW system point designs

    International Nuclear Information System (INIS)

    Yang, W.S.; Taiwo, T.A.; Hill, R.N.; Khalil, H.S.; Wade, D.C.

    2001-01-01

    As part of the Advanced Accelerator Application (AAA) program in the U.S., preliminary design studies have been performed at Argonne National Laboratory (ANL) and Los Alamos National Laboratory (LANL) to define and compare candidate Accelerator Transmutation of Waste (ATW) systems. The studies at ANL have focused primarily on the transmutation blanket component of the overall system. Lead-bismuth eutectic (LBE), sodium, and gas cooled systems are among the blanket technology options currently under consideration. This paper summarizes the results from neutronics trade studies performed at ANL. Core designs have been developed for LBE and sodium cooled 840 MWt fast spectrum accelerator driven systems employing re-cycle. Additionally, neutronics analyses have been performed for a helium-cooled 600 MWt hybrid thermal and fast spectrum system proposed by General Atomics (GA), which is operated in the critical mode for three cycles and in a subcritical accelerator driven mode for a subsequent single cycle. For these three point designs, isotopic inventories, consumption rates, and annual burnup rates are compared. The mass flows and the ultimate loss of transuranic (TRU) isotopes to the waste stream per unit of heat generated during transmutation are also compared on a consistent basis. (author)

  6. Utilization of multi-purpose high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kawada, Osamu; Onuki, Yoshiaki; Wasaoka, Takeshi.

    1974-01-01

    Concerning the utilization of multi-purpose high temperature gas-cooled reactors, the electric power generation with gas turbines is described: features of HTR-He gas turbine power plants; the state of development of He gas turbines; and combined cycle with gas turbines and steam turbines. The features of gas turbines concern heat dissipation into the environment and the mode of load operation. Outstanding work in the development of He gas turbines is that in Hochtemperatur Helium-Turbine Project in West Germany. The power generation with combined gas turbines and steam turbines appears to be superior to that with gas turbines alone. (Mori, K.)

  7. Theoretical thermodynamics analysis of cooling cycle bu advanced gas absorption using solar energy; Analisis teorico-experimental de un ciclo de refrigeracion por absorcion avanzado gax, operando con energia solar

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, V. E.; Vidal, A. S.; Garcia, C. A.; Garcia-Valladares, O.; Best, R. B.; Hernandez, J. G.; Velazquez, N. L.

    2004-07-01

    In this article a solar system of refrigeration by absorption with heat exchange generator absorber (GAX) was analyzed. A theoretical thermodynamic analysis of the energetic behavior of the GAX absorption system was made. Experimental results were obtained with generation temperatures of 190 and 220 C, the evaporation temperature was set at 9 C and temperatures of cooling fluids (air and water) were set at 30 C and 28 C, respectively. It was possible to appreciate that the GAX effect decrease whether absorber, type falling film, is operated in option of parallel flow and it was increased when the absorber was operated in option of counterflow. (Author)

  8. Simulation of a gas cooled reactor with the system code CATHARE

    International Nuclear Information System (INIS)

    Bentivoglio, Fabrice; Ruby, Alain; Geffraye, Genevieve; Messie, Anne; Saez, Manuel; Tauveron, Nicolas; Widlund, Ola

    2006-01-01

    In recent years the CEA has commissioned a wide range of feasibility studies of future advanced nuclear reactors, in particular gas-cooled reactors (GCR). This paper presents an overview of the use of the thermohydraulics code CATHARE in these activities. Extensively validated and qualified for pressurized water reactors, CATHARE has been adapted to deal also with gas-cooled reactor applications. Rather than branching off a separate GCR version of CATHARE, new features have been integrated as independent options in the standard version of the code, respecting the same stringent procedures for documentation and maintenance. CATHARE has evolved into an efficient tool for GCR applications, with first results in good agreement with existing experimental data and other codes. The paper give an example among the studies already carried out with CATHARE with the case of the Very High Temperature Reactor (VHTR) concepts. Current and future activities for experimental validation of CATHARE for GCR applications are also discussed. Short-term validation activities are also included with the assessment of the German utility Oberhausen II. For the long term, CEA has initiated an ambitious experimental program ranging from small scale loops for physical correlations to component technology and system demonstration loops. (authors)

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

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-01-01

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

  10. Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Wang Yan; Li Fu; Zheng Yanhua

    2014-01-01

    In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)

  11. Demonstration of Enabling Spar-Shell Cooling Technology in Gas Turbines

    Energy Technology Data Exchange (ETDEWEB)

    Downs, James [Florida Turbine Technologies Inc., Jupiter, FL (United States)

    2014-12-29

    In this Advanced Turbine Program-funded Phase III project, Florida Turbine Technologies, Inc. (FTT) has developed and tested, at a pre-commercial prototypescale, spar-shell turbine airfoils in a commercial gas turbine. The airfoil development is based upon FTT’s research and development to date in Phases I and II of Small Business Innovative Research (SBIR) grants. During this program, FTT has partnered with an Original Equipment Manufacturer (OEM), Siemens Energy, to produce sparshell turbine components for the first pre-commercial prototype test in an F-Class industrial gas turbine engine and has successfully completed validation testing. This project will further the commercialization of this new technology in F-frame and other highly cooled turbine airfoil applications. FTT, in cooperation with Siemens, intends to offer the spar-shell vane as a first-tier supplier for retrofit applications and new large frame industrial gas turbines. The market for the spar-shell vane for these machines is huge. According to Forecast International, 3,211 new gas turbines units (in the >50MW capacity size range) will be ordered in ten years from 2007 to 2016. FTT intends to enter the market in a low rate initial production. After one year of successful extended use, FTT will quickly ramp up production and sales, with a target to capture 1% of the market within the first year and 10% within 5 years (2020).

  12. AGT101 Advanced Gas Turbine Technology update

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, G.L.; Kidwell, J.R.; Kreiner, D.M.

    1986-01-01

    The Garrett/Ford Advanced Gas Turbine Technology Development Program, designated AGT101, has made significant progress during 1985 encompassing ceramic engine and ceramic component testing. Engine testing has included full speed operation to 100,000 rpm and 1149C (2100F) turbine inlet temperature, initial baseline performance mapping and ceramic combustor start and steady state operation. Over 380 hours of test time have been accumulated on four development engines. High temperature foil bearing coatings have passed rig test and a thick precious metal foil coating selected for engine evaluation. Ceramic structures have been successfully rig tested at 1371C (2500F) for over 27 hours.

  13. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  14. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh; Cliff Davis; Goon C. Park

    2007-01-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls

  15. Gas-fired cogeneration and cooling: new study identifies major benefits

    International Nuclear Information System (INIS)

    Watt, G.

    2001-01-01

    A research paper- 'Gas Fired Cogeneration and Cooling: Markets, Technologies and Greenhouse Gas Savings'- launched at last month's Australian Gas Association 2001 Convention, reveals that gas cooling could replace 25 PJ of electricity summer demand, and reduce greenhouse gas emissions by 58 percent compared with electrical technologies. Commissioned by the AGA's Gas Cooling Task Force and supported by the Sustainable Energy Authority of Victoria and the Sustainable Energy Development Authority of NSW, the study examined market opportunities and environmental outcomes for the combined gas cogeneration and cooling technologies. It shows that the penetration of gas into the distributed cooling and power generation market is being driven by the following developments: the uncertainty and volatility of electricity costs, particularly during summer, electricity market structural changes which encourage distributed generation, high and uncertain world oil prices, the relative stability of Australian gas prices, the encouragement of demand and energy management strategies by regulators, greenhouse gas emission reduction policies, indoor air quality issues, product and productivity improvements in industry and CFC phase-out opportunities

  16. Proposal of a fast gas-cooled reactor using transuranics

    International Nuclear Information System (INIS)

    Macedo, Anderson Altair Pinheiro de

    2016-01-01

    In the last two decades, nations that have invested in research and energy generation through nuclear source have devoted part of their efforts in developing new technologies for nuclear reactors. Part of this investment focuses on new material testing, particularly regarding new fuels. In a world view that breaths sustainability, the reprocess and reuse of spent fuel from conventional reactors comes alive in nuclear technology, presenting itself as a real alternative of energy source for the latest generation of reactors. Different concepts of fourth generation reactors have been proposed and must meet some basic requirements, such as: extended burnup, improvement of passive safety, better radioactive waste management, possibility to use reprocessed fuel and proliferation resistance. In this context, the GFR (Gas-cooled Fast Reactor) is one of the future promises, presenting satisfactory neutronic results on the use of type of fuel (U, Pu) C. In the present work, the fuel of a traditional GFR reactor that uses (U, Pu)C was sub was replaced by a transuranic reprocessed fuel (TRU), obtained by non-proliferation reprocessing technology. The UO 2 fuel initially enriched by 3.1% was burned in a standard PWR, with full burn of 33,000 MWd/T. Afterward it was left in a pool for 5 years and finally reprocessed by UREX + method. Two fuels were studied and evaluated, one diluted with depleted uranium (U, TRU)C, and the other diluted in thorium (Th, TRU)C. Assessments were done in steady state and as well as during burning and were compared with results obtained using the standard fuel, (U, Pu) C. The outcome shows that the use of TRU as a fuel, in GFR type reactors, is a real possibility. The research was done using the SCALE 6.0 code modules. (author)

  17. Improved spacers for high temperature gas-cooled heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Nordstroem, L A [Swiss Federal Institute for Reactor Research, Wuerenlingen (Switzerland)

    1984-07-01

    Experimental and analytical investigations in the field of heat exchanger thermohydraulics have been performed at EIR for many years, Basic studies have been carried out on heat transfer and pressure loss for tube bundles of different geometries and tube surfaces. As a part of this overall R+D programme for heat exchangers, investigations have been carried out on spacer pressure loss in bundles with longitudinal flow. An analytical spacer pressure loss model was developed which could handle different types of subchannel within the bundle. The model has been evaluated against experiments, using about 25 spacers of widely differing geometries. In a gas-cooled reactor it is important to keep the pressure loss over the primary circuit heat exchangers to a minimum. In exchangers with grid spacers these contribute a significant proportion of the overall bundle losses. For example, in the HHT Recuperator, with a shell-side pressure loss of 3.5 % of the inlet pressure, the spacers cause about one half of this loss. Reducing the loss to, say, 2.5 % results in an overall increase in plant efficiency by more than 1 % - a significant improvement Preliminary analysis identified 5 geometries in particular which were chosen for experimental evaluation as part of a joint project with the SULZER Company, to develop a low pressure-loss spacer for HHT heat exchangers (longitudinal counter-flow He/He and He/H{sub 2}O designs). The aim of the tests was to verify the low pressure-loss characteristics of these spacer grid types, as well as the quality of the results calculated by the computer code analytical model. The experimental and analytical results are compared in this report.

  18. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, October 1, 1982-March 3, 1983

    International Nuclear Information System (INIS)

    1983-06-01

    This report provides descriptions and results of the technical effort during the first half of FY 83 on the Gas-Cooled Thermal Reactor Program. The work on Integration and Management (WBS 01) includes the preparation of the Advanced Systems Concept Evaluation Plan and the Advanced Systems Technology Development Plan in addition to the program management activities. The Market Definition (WBS 03) efforts considered the application of the Modular Reactor System with reforming (MRS-R) to the production of methanol and ammonia and the refining of petroleum. Within the Plant Technology (WBS 13) task there were activities to develop anlytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. In addition to the work on the advanced HTGR for process heat users, new activities were initiated in support of the HTGR-SC/C Lead plant Protect (WBS 30 and 31). The Plant Simulation task (WBS 31) was initiated to develop a computer code for simulation of plant operation and for plant transient systems analysis. The efforts on the advanced HTGR systems was performed under the Modular Systems task (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors

  19. Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    International Nuclear Information System (INIS)

    Silady, F.A.; Millunzi, A.C.

    1989-08-01

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab

  20. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS's heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis

  1. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  2. The role of the IAEA in gas-cooled reactor development and application

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, J.; Kupitz, J. [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power

    1995-03-01

    The IAEA`s activities in gas-cooled reactor development focus on the four technical areas which are predicted to provide advanced HTGRs with a high degree of safety, but which must be proven. These are: (a) the safe neutronic behaviour of the core, (b) reliance on ceramic coated fuel particles to retain fission products even under extreme conditions, (c) the ability to dissipate decay heat by natural heat transport mechanisms, and (d) the safe behaviour of the fuel and reactor core under chemical attack (air or water ingress). The first three are subjects of Coordinated Research Programmes (CRPs) and the last was recently addressed in an information exchange meeting. CRPs are 3 to 6 years in duration, and often involve experimental activities. CRPs allow a sharing of efforts on an international basis and benefit from the experience and expertise of researchers from the participating institutes. (J.P.N.).

  3. A study of silver behavior in Gas-turbine High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Tanaka, Toshiyuki

    1995-11-01

    A Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR) is one of the promising reactor systems of future HTGRs. In the design of GT-HTGR, behavior of fission products, especially of silver, is considered to be important from the view point of maintenance of gas-turbine. A study of silver behavior in the GT-HTGR was carried out based on current knowledge. The purposes of this study were to determine an importance of the silver problem quantitatively, countermeasures to the problem and items of future research and development which will be needed. In this study, inventory, fractional release from fuel, plateout in the primary circuit and radiation dose were evaluated, respectively. Based on this study, it is predicted that gamma-ray from plateout silver in gas-turbine system contributes about a half of total radiation dose after reactor shutdown. In future, more detail data for silver release from fuel, plateout behavior, etc. using the High Temperature Engineering Test Reactor (HTTR), for example, will be needed to carry out reasonable design. (author)

  4. Detection of gas-permeable fuel particles for highl 7490 temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Thiele, B.A.; Stinton, D.P.; Costanzo, D.A.

    1980-01-01

    Fuel for High-Temperature Gas-Cooled Reactors (HTGR) consists of uranium oxide-carbide and thoria microspheres coated with layers of pyrolytic carbon and silicon carbide. The pyrolytic carbon coatings must be gas-tight to perform properly during irradiation. Therefore, particles must be carefully characterized to determine the number of defective particles (ie bare kernels, and cracked or permeable coatings). Although techniques are available to determine the number of bare kernels or cracked coatings, no reliable technique has been available to measure coating permeability. This work describes a technique recently developed to determine whether coatings for a batch of particles are gas-tight or permeable. Although most of this study was performed on Biso-coated particles, the technique applies equally well to Triso-coated particles. About 150 randomly selected Biso-particle batches were studied in this work. These batches were first subjected to an 18-hr chlorination at 15000C, and the volatile thorium tetrachloride released through cracked or very permeable coatings was measured versus chlorination time. Chlorinated batches were also radiographed to detect any thorium that had migrated from the kernel into the coatings. From this work a technique was developed to determine coating permeability. This consists of an 18-hr chlorination of multiple samples without measurement of the heavy metal released. Each batch is then radiographed and the heavy metal diffusion within each particle is examined so it can be determined if a particle batch is permeable, slightly permeable, or gas-tight. (author)

  5. Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

    International Nuclear Information System (INIS)

    Badulescu, A.; Groeneveld, D.C.

    2000-01-01

    Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

  6. Advanced adsorption cooling cum desalination cycle: A thermodynamic framework

    KAUST Repository

    Chakraborty, Anutosh; Thu, Kyaw; Ng, K. C.

    2011-01-01

    We have developed a thermodynamic framework to calculate adsorption cooling cum desalination cycle performances as a function of pore widths and pore volumes of highly porous adsorbents, which are formulated from the rigor of thermodynamic property

  7. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  8. Coated particle fuel for high temperature gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl; Nabielek, Heinz [Research Center Julich (FZJ), Julich (Germany); Kendall, James M. [Global Virtual L1c, Prescott (United States)

    2007-10-15

    applications at 850-900 .deg. C and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 {mu}m diameter UO{sub 2} kernel of 10% enrichment is surrounded by a 100 {mu}m thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 {mu}m thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level.

  9. Coated particle fuel for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Verfondern, Karl; Nabielek, Heinz; Kendall, James M.

    2007-01-01

    and for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 μm diameter UO 2 kernel of 10% enrichment is surrounded by a 100 μm thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 μm thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level

  10. Economizer Based Data Center Liquid Cooling with Advanced Metal Interfaces

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Chainer

    2012-11-30

    A new chiller-less data center liquid cooling system utilizing the outside air environment has been shown to achieve up to 90% reduction in cooling energy compared to traditional chiller based data center cooling systems. The system removes heat from Volume servers inside a Sealed Rack and transports the heat using a liquid loop to an Outdoor Heat Exchanger which rejects the heat to the outdoor ambient environment. The servers in the rack are cooled using a hybrid cooling system by removing the majority of the heat generated by the processors and memory by direct thermal conduction using coldplates and the heat generated by the remaining components using forced air convection to an air- to- liquid heat exchanger inside the Sealed Rack. The anticipated benefits of such energy-centric configurations are significant energy savings at the data center level. When compared to a traditional 10 MW data center, which typically uses 25% of its total data center energy consumption for cooling this technology could potentially enable a cost savings of up to $800,000-$2,200,000/year (assuming electricity costs of 4 to 11 cents per kilowatt-hour) through the reduction in electrical energy usage.

  11. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  12. Influence of precooling cooling air on the performance of a gas turbine combined cycle

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Ik Hwan; Kang, Do Won; Kang, Soo Young; Kim, Tong Seop [Inha Univ., Incheon (Korea, Republic of)

    2012-02-15

    Cooling of hot sections, especially the turbine nozzle and rotor blades, has a significant impact on gas turbine performance. In this study, the influence of precooling of the cooling air on the performance of gas turbines and their combined cycle plants was investigated. A state of the art F class gas turbine was selected, and its design performance was deliberately simulated using detailed component models including turbine blade cooling. Off design analysis was used to simulate changes in the operating conditions and performance of the gas turbines due to precooling of the cooling air. Thermodynamic and aerodynamic models were used to simulate the performance of the cooled nozzle and rotor blade. In the combined cycle plant, the heat rejected from the cooling air was recovered at the bottoming steam cycle to optimize the overall plant performance. With a 200K decrease of all cooling air stream, an almost 1.78% power upgrade due to increase in main gas flow and a 0.70 percent point efficiency decrease due to the fuel flow increase to maintain design turbine inlet temperature were predicted.

  13. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

    2014-01-01

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

  14. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    1985-01-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  15. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity.

  16. Analytical Modelling of the Effects of Different Gas Turbine Cooling Techniques on Engine Performance =

    Science.gov (United States)

    Uysal, Selcuk Can

    In this research, MATLAB SimulinkRTM was used to develop a cooled engine model for industrial gas turbines and aero-engines. The model consists of uncooled on-design, mean-line turbomachinery design and a cooled off-design analysis in order to evaluate the engine performance parameters by using operating conditions, polytropic efficiencies, material information and cooling system details. The cooling analysis algorithm involves a 2nd law analysis to calculate losses from the cooling technique applied. The model is used in a sensitivity analysis that evaluates the impacts of variations in metal Biot number, thermal barrier coating Biot number, film cooling effectiveness, internal cooling effectiveness and maximum allowable blade temperature on main engine performance parameters of aero and industrial gas turbine engines. The model is subsequently used to analyze the relative performance impact of employing Anti-Vortex Film Cooling holes (AVH) by means of data obtained for these holes by Detached Eddy Simulation-CFD Techniques that are valid for engine-like turbulence intensity conditions. Cooled blade configurations with AVH and other different external cooling techniques were used in a performance comparison study. (Abstract shortened by ProQuest.).

  17. The variation of particle gas-borne concentration with time in a gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reed, J; Hall, D; Reeks, M W [Central Electricity Generating Board, Berkeley Nuclear Laboratories (United Kingdom)

    1985-07-01

    If volatile fission products are released from fuel during a reactor fault, a significant fraction could become attached to small particles also present in the coolant. In such circumstances the retention of those particles by the reactor circuit will limit the level of gas-borne particle concentration and hence be important in reducing the potential release of fission product activity to the atmosphere. Clearly the retention of particles will be influenced by both the deposition and resuspension of particles from surfaces exposed to the coolant flow. In this paper we consider deposition and resuspension but pay particular attention to the role of resuspension, which in the past has been given little consideration. A recently developed model for the resuspension of small particles by a turbulent flow is outlined. Traditionally, resuspension has been interpreted as a force balance between the aerodynamic removal forces and the surface adhesive forces. In contrast, this new approach embodies an energy balance criterion for particle resuspension. Furthermore, the stochastic nature of this new model has shown that resuspension can be sub-divided into two regimes: (i) initial resuspension (resuspension occurring in times less than a second) which reduces the net deposition of particles to a surface; and (ii) longer term resuspension (resuspension after 1 second) which determines the asymptotic decay of particle gas-borne concentration. It is seen that the asymptotic decay varies almost inversely as the decay time. Force balance models are unsuccessful in accounting for the experimentally observed longer term resuspension. We show that a Volterra integro-differential equation best describes the variation of particle gas-borne concentration with time in a recirculating gas flow such as a gas cooled reactor. It is seen that the longer term resuspension has a major influence in the final decay of particle concentration. (author)

  18. The variation of particle gas-borne concentration with time in a gas cooled reactor

    International Nuclear Information System (INIS)

    Reed, J.; Hall, D.; Reeks, M.W.

    1985-01-01

    If volatile fission products are released from fuel during a reactor fault, a significant fraction could become attached to small particles also present in the coolant. In such circumstances the retention of those particles by the reactor circuit will limit the level of gas-borne particle concentration and hence be important in reducing the potential release of fission product activity to the atmosphere. Clearly the retention of particles will be influenced by both the deposition and resuspension of particles from surfaces exposed to the coolant flow. In this paper we consider deposition and resuspension but pay particular attention to the role of resuspension, which in the past has been given little consideration. A recently developed model for the resuspension of small particles by a turbulent flow is outlined. Traditionally, resuspension has been interpreted as a force balance between the aerodynamic removal forces and the surface adhesive forces. In contrast, this new approach embodies an energy balance criterion for particle resuspension. Furthermore, the stochastic nature of this new model has shown that resuspension can be sub-divided into two regimes: (i) initial resuspension (resuspension occurring in times less than a second) which reduces the net deposition of particles to a surface; and (ii) longer term resuspension (resuspension after 1 second) which determines the asymptotic decay of particle gas-borne concentration. It is seen that the asymptotic decay varies almost inversely as the decay time. Force balance models are unsuccessful in accounting for the experimentally observed longer term resuspension. We show that a Volterra integro-differential equation best describes the variation of particle gas-borne concentration with time in a recirculating gas flow such as a gas cooled reactor. It is seen that the longer term resuspension has a major influence in the final decay of particle concentration. (author)

  19. AGT 101 - Advanced Gas Turbine technology update

    Energy Technology Data Exchange (ETDEWEB)

    Kidwell, J.R.; Kreiner, D.M.

    1985-03-01

    The Advanced Gas Turbine (AGT) 101 program has made significant progress during 1984 in ceramic component and engine test bed development, including initial ceramic engine testing. All ceramic components for the AGT 101 (1644 K) engine are now undergoing development. Ceramic structures have been undergoing extensive analysis, design modification, and rig testing. AGT 101 (1644 K) start capability has been demonstrated in rig tests. Also, 1644 K steady-state testing has been initiated in the test rigs to obtain a better understanding of ceramics in that environment. The ceramic turbine rotor has progressed through cold spin test 12,040 rad/sec and hot turbine rig test, and is currently in initial phases of engine test. Over 400 hours of engine testing is expected by March 1985, including approximately 150 hours of operation and 50 starts on the 1422 K engine. All activities are progressing toward 1644 K engine testing in mid-1985.

  20. ADVANCED TECHNOLOGIES FOR STRIPPER GAS WELL ENHANCEMENT

    International Nuclear Information System (INIS)

    Charles M. Boyer II; Ronald J. MacDonald P.G.

    2001-01-01

    As part of Task 1 in Advanced Technologies for Stripper Gas Well Enhancement, Schlumberger-Holditch Reservoir Technologies (H-RT) has joined with two Appalachian Basin producers, Great Lakes Energy Partners, LLC, and Belden and Blake Corporation to develop methodologies for identification and enhancement of stripper wells with economic upside potential. These industry partners have provided us with data for more than 700 wells in northwestern Pennsylvania. Phase 1 goals of this project are to develop and validate methodologies that can quickly and cost-effectively identify wells with enhancement potential. We are currently in the final stages of developing and testing our new Microsoft(trademark) Access/Excel based software. We will be processing this well data and identifying potential candidate wells that can be used in Phase 2 to validate these methodologies. Preparation of the final technical report is underway

  1. ADVANCED TECHNOLOGIES FOR STRIPPER GAS WELL ENHANCEMENT

    International Nuclear Information System (INIS)

    Charles M. Boyer II; Ronald J. MacDonald P.G.

    2001-01-01

    As part of Task 1 in Advanced Technologies for Stripper Gas Well Enhancement, Schlumberger-Holditch Reservoir Technologies (H-RT) has joined with two Appalachian Basin producers, Great Lakes Energy Partners, LLC, and Belden and Blake Corporation to develop methodologies for identification and enhancement of stripper wells with economic upside potential. These industry partners have provided us with data for more than 700 wells in northwestern Pennsylvania. Phase 1 goals of this project are to develop and validate methodologies that can quickly and cost-effectively identify wells with enhancement potential. We have continued to enhance and streamline our software, and we are testing the final stages of our new Microsoft(trademark) Access/Excel based software. We are continuing to process this well data and are identifying potential candidate wells that can be used in Phase 2 to validate the new methodologies. In addition, preparation of the final technical report is underway

  2. Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Jie; Gu Yihua

    2005-01-01

    The high temperature gas-cooled reactor (HTGR) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the various types of HTGR gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. Furthermore they are theoretically converted to three Brayton cycles of helium, nitrogen and air. Those three types of Brayton cycles are thermodynamically analyzed and optimized. The results show that the variety of gas affects the cycle pressure ratio more significantly than other cycle parameters, however, the optimized cycle efficiencies of the three Brayton cycles are almost the same. In addition, the turbomachines which are required for the three optimized Brayton cycles are aerodynamically analyzed and compared and their fundamental characteristics are obtained. Helium turbocompressor has lower stage pressure ratio and more stage number than those for nitrogen and air machines, while helium and nitrogen turbocompressors have shorter blade length than that for air machine

  3. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    International Nuclear Information System (INIS)

    Weaver, K. D.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Weaver, K.D.

    2005-01-01

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

  5. Hot gas path component cooling system having a particle collection chamber

    Science.gov (United States)

    Miranda, Carlos Miguel; Lacy, Benjamin Paul

    2018-02-20

    A cooling system for a hot gas path component includes a substrate having an outer surface and an inner surface. The inner surface defines at least one interior space. A passage is formed in the substrate between the outer surface and the inner surface. An access passage is formed in the substrate and extends from the outer surface to the inner space. The access passage is formed at a first acute angle to the passage and includes a particle collection chamber. The access passage is configured to channel a cooling fluid to the passage. Furthermore, the passage is configured to channel the cooling fluid therethrough to cool the substrate.

  6. Indiana State University Graduates to Advanced Plastic Cooling Towers

    Science.gov (United States)

    Sullivan, Ed

    2012-01-01

    Perhaps more than many other industries, today's universities and colleges are beset by dramatically rising costs on every front. One of the areas where overhead can be contained or reduced is in the operation of the chilled water systems that support air conditioning throughout college campuses, specifically the cooling towers. Like many…

  7. Engineered Materials for Advanced Gas Turbine Engine, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This project will develop innovative composite powders and composites that will surpass the properties of currently identified materials for advanced gas turbine...

  8. Gas cooled reactor decommissioning. Packaging of waste for disposal in the United Kingdom deep repository

    International Nuclear Information System (INIS)

    Barlow, S.V.; Wisbey, S.J.; Wood, P.

    1998-01-01

    United Kingdom Nirex Limited has been established to develop and operate a deep underground repository for the disposal of the UK's intermediate and certain low level radioactive waste. The UK has a significant Gas Cooled Reactor (GCR) programme, including both Magnox and AGR (Advanced Gas-cooled Reactor) capacity, amounting to 26 Magnox reactors, 15 AGR reactors as well as research and prototype reactor units such as the Windscale AGR and the Windscale Piles. Some of these units are already undergoing decommissioning and Nirex has estimated that some 15,000 m 3 (conditioned volume) will come forward for disposal from GCR decommissioning before 2060. This volume does not include final stage (Stage 3) decommissioning arisings from commercial reactors since the generating utilities in the UK are proposing to adopt a deferred safe store strategy for these units. Intermediate level wastes arising from GCR decommissioning needs to be packaged in a form suitable for on-site interim storage and eventual deep disposal in the planned repository. In the absence of Conditions for Acceptance for a repository in the UK, the dimensions, key features and minimum performance requirements for waste packages are defined in Waste Package Specifications. These form the basis for all assessments of the suitability of wastes for disposal, including GCR wastes. This paper will describe the nature and characteristics of GCR decommissioning wastes which are intended for disposal in a UK repository. The Nirex Waste Package Specifications and the key technical issues, which have been identified when considering GCR decommissioning waste against the performance requirements within the specifications, are discussed. (author)

  9. Development of cooling and cleaning systems for enhanced gas ...

    African Journals Online (AJOL)

    In order to address these tar related problems a cleaning and cooling system has been developed in house that facilitates tar removal to acceptable levels tolerated by the internal combustion (IC) engine and meets emission standards as well. The main objective of the present work is to reduce tar level and develop control ...

  10. Noble gas, binary mixtures for commercial gas-cooled reactor systems

    International Nuclear Information System (INIS)

    El-Genk, M. S.; Tournier, J. M.

    2007-01-01

    Commercial gas cooled reactors employ helium as a coolant and working fluid for the Closed Brayton Cycle (CBC) turbo-machines. Helium has the highest thermal conductivity and lowest dynamic viscosity of all noble gases. This paper compares the relative performance of pure helium to binary mixtures of helium and other noble gases of higher molecular weights. The comparison is for the same molecular flow rate, and same operating temperatures and geometry. Results show that although helium is a good working fluid because of its high heat transfer coefficient and significantly lower pumping requirement, a binary gas mixture of He-Xe with M = 15 gm/mole has a heat transfer coefficient that is ∼7% higher than that of helium and requires only 25% of the number stages of the turbo-machines. The binary mixture, however, requires 3.5 times the pumping requirement with helium. The second best working fluid is He-Kr binary mixture with M = 10 gm/mole. It has 4% higher heat transfer coefficient than He and requires 30% of the number of stages in the turbo-machines, but requires twice the pumping power

  11. Advances in electron cooling in heavy-ion storage rings

    International Nuclear Information System (INIS)

    Danared, H.

    1994-01-01

    The efficiency of electron cooling can be improved by reducing the temperature of the electrons. If the magnetic field at the location of the electron gun is stronger than in the region where the electrons interact with the ions, and the field gradient is adiabatic with respect to the cyclotron motion of the electrons, the resulting expansion of the electron beam reduces its transverse temperature by a factor equal to the ratio between the two fields. A ten times expanded electron beam was introduced in the CRYRING electron cooler in the summer of 1993, and similar arrangements have since then been made at the TSR ring in Heidelberg and at ASTRID in Aarhus. The reduction of the transverse electron temperature has increased cooling rates with large factors, and improves the energy resolution and increases count rates when the cooler is used as an electron target for ion-electron recombination experiments

  12. Advanced adsorption cooling cum desalination cycle: A thermodynamic framework

    KAUST Repository

    Chakraborty, Anutosh

    2011-01-01

    We have developed a thermodynamic framework to calculate adsorption cooling cum desalination cycle performances as a function of pore widths and pore volumes of highly porous adsorbents, which are formulated from the rigor of thermodynamic property surfaces of adsorbent-adsorbate system and the adsorption interaction potential between them. Employing the proposed formulations, the coefficient of performance (COP) and overall performance ratio (OPR) of adsorption cycle are computed for various pore widths of solid adsorbents. These results are compared with experimental data for verifying the proposed thermodynamic formulations. It is found from the present analysis that the COP and OPR of adsorption cooling cum desalination cycle is influenced by (i) the physical characteristics of adsorbents, (ii) characteristics energy and (iii) the surface-structural heterogeneity factor of adsorbent-water system. The present study confirms that there exists a special type of adsorbents having optimal physical characteristics that allows us to obtain the best performance.

  13. Recombining processes in a cooling plasma by mixing of initially heated gas

    International Nuclear Information System (INIS)

    Furukane, Utaro; Sato, Kuninori; Takiyama, Ken; Oda, Toshiatsu.

    1992-03-01

    A numerical investigation of recombining process in a high temperature plasma in a quasi-steady state is made in a gas contact cooling, in which the initial temperature effect of contact gas heated up by the hot plasma is considered as well as the gas cooling due to the surrounding neutral particles freely coming into the plasma. The calculation has shown that the electron temperature relaxes in accord with experimental results and that the occurrence of recombining region and the inverted populations almost agree with the experimental ones. (author)

  14. Advanced turbine cooling, heat transfer, and aerodynamic studies

    Energy Technology Data Exchange (ETDEWEB)

    Je-Chin Han; Schobeiri, M.T. [Texas A& M Univ., College Station, TX (United States)

    1995-10-01

    The contractual work is in three parts: Part I - Effect of rotation on enhanced cooling passage heat transfer, Part II - Effect on Thermal Barrier Coating (TBC) spallation on surface heat transfer, and Part III - Effect of surface roughness and trailing edge ejection on turbine efficiency under unsteady flow conditions. Each section of this paper has been divided into three parts to individually accommodate each part. Part III is further divided into Parts IIIa and IIIb.

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

    Science.gov (United States)

    Palmer, David J.

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

  16. Nuclear powerplant with closed gas-cooling circuit

    International Nuclear Information System (INIS)

    Haferkamp, D.; Hodzic, A.; Winter, U.

    1976-01-01

    Disclosed is a nuclear power plant comprising a pressure-tight safety vessel surrounding the entire plant, an inner vessel of reinforced concrete, a high-temperature reactor contained in the inner vessel, a gas turbine assembly having a turbine and a high- and low-pressure compressor located in a horizontally oriented chamber below the reactor, a plurality of heat exchange units positioned in a plurality of vertically oriented pods spaced radially symmetrically about the reactor and suitable conduits for carrying the reactive gas between the system components. The conduits are arranged in generally horizontally and vertically oriented straight lines, and the conduits for carrying low-pressure gas comprise a horizontal system positioned beneath the turbine assembly having a plurality of coaxial connecting tubes, collectors and distributors as well as normal conduits, so that high pressure gas flows in the internal passage and low-pressure gas flows in the outer passage. 22 claims, 7 figures

  17. Hot gas path component having near wall cooling features

    Science.gov (United States)

    Miranda, Carlos Miguel; Kottilingam, Srikanth Chandrudu; Lacy, Benjamin Paul

    2017-11-28

    A method for providing micro-channels in a hot gas path component includes forming a first micro-channel in an exterior surface of a substrate of the hot gas path component. A second micro-channel is formed in the exterior surface of the hot gas path component such that it is separated from the first micro-channel by a surface gap having a first width. The method also includes disposing a braze sheet onto the exterior surface of the hot gas path component such that the braze sheet covers at least of portion of the first and second micro-channels, and heating the braze sheet to bond it to at least a portion of the exterior surface of the hot gas path component.

  18. Performance and economic enhancement of cogeneration gas turbines through compressor inlet air cooling

    Science.gov (United States)

    Delucia, M.; Bronconi, R.; Carnevale, E.

    1994-04-01

    Gas turbine air cooling systems serve to raise performance to peak power levels during the hot months when high atmospheric temperatures cause reductions in net power output. This work describes the technical and economic advantages of providing a compressor inlet air cooling system to increase the gas turbine's power rating and reduce its heat rate. The pros and cons of state-of-the-art cooling technologies, i.e., absorption and compression refrigeration, with and without thermal energy storage, were examined in order to select the most suitable cooling solution. Heavy-duty gas turbine cogeneration systems with and without absorption units were modeled, as well as various industrial sectors, i.e., paper and pulp, pharmaceuticals, food processing, textiles, tanning, and building materials. The ambient temperature variations were modeled so the effects of climate could be accounted for in the simulation. The results validated the advantages of gas turbine cogeneration with absorption air cooling as compared to other systems without air cooling.

  19. MHD/gas turbine systems designed for low cooling water requirements

    International Nuclear Information System (INIS)

    Annen, K.D.; Eustis, R.H.

    1983-01-01

    The MHD/gas turbine combined-cycle system has been designed specifically for applications where the availability of cooling water is very limited. The base case systems which were studied consist of a coal-fired MHD plant with an air turbine bottoming plant and require no cooling water. In addition to the base case systems, systems were considered which included the addition of a vapor cycle bottoming plant to improve the thermal efficiency. These systems require a small amount of cooling water. The results show that the MHD/gas turbine systems have very good thermal and economic performances. The base case I MHD/gas turbine system (782 MW /SUB e/ ) requires no cooling water, has a heat rate which is 13% higher, and a cost of electricity which is only 7% higher than a comparable MHD/steam system (878 MW /SUB e/ ) having a cooling tower heat load of 720 MW. The case I vapor cycle bottomed systems have thermal and economic performances which approach and even exceed those of the MHD/steam system, while having substantially lower cooling water requirements. Performances of a second-generation MHD/gas turbine system and an oxygen-enriched, early commercial system are also evaluated. An analysis of nitric oxide emissions shows compliance with emission standards

  20. ADVANCED TECHNOLOGIES FOR STRIPPER GAS WELL ENHANCEMENT

    International Nuclear Information System (INIS)

    Charles M. Boyer II; Ronald J. MacDonald P.G.

    2002-01-01

    As part of Task 1 in Advanced Technologies for Stripper Gas Well Enhancement, Schlumberger-Holditch Reservoir Technologies (H-RT) has joined with two Appalachian Basin producers, Great Lakes Energy Partners, LLC, and Belden and Blake Corporation to develop methodologies for identification and enhancement of stripper wells with economic upside potential. These industry partners have provided us with data for more than 700 wells in northwestern Pennsylvania. Phase 1 goals of this project are to develop and validate methodologies that can quickly and cost-effectively identify wells with enhancement potential. We have continued to enhance and streamline our software, and we are testing the final stages of our new Microsoft(trademark) Access/Excel based software. We are continuing to process the information and are identifying potential candidate wells that can be used in Phase 2 to validate the new methodologies. In addition, preparation of the final technical report is underway. During this quarter, we have presented our project and discussed the software to numerous Petroleum Technology Transfer Council (PTTC) workshops located in various regions of the United States

  1. Low grade heat driven adsorption system for cooling and power generation using advanced adsorbent materials

    International Nuclear Information System (INIS)

    Al-Mousawi, Fadhel Noraldeen; Al-Dadah, Raya; Mahmoud, Saad

    2016-01-01

    Highlights: • Adsorption system based on water and advanced physical adsorbents has the potential of producing cooling and power. • Adding an expander to physisorption system enhances efficiency by up to 11%. • MIL101Cr MOF can produce 95 W/kg and 1357 W/kg of specific power and cooling. • AQSOA Z02 can produce 73 W/kg and 640 W/kg of specific power and cooling. - Abstract: Globally there is abundance of low grade heat sources (around 150 °C) from renewables like solar energy or from industrial waste heat. The exploitation of such low grade heat sources will reduce fossil fuel consumption and CO_2 emissions. Adsorption technology offers the potential of using such low grade heat to generate cooling and power. In this work, the effect of using advanced adsorbent materials like AQSOA-Z02 (SAPO-34) zeolite and MIL101Cr Metal Organic Framework (MOF) at various operating conditions on power and cooling performance compared to that of commonly used silica-gel was investigated using water as refrigerant. A mathematical model for a two bed adsorption cooling cycle has been developed with the cycle modified to produce power by incorporating an expander between the desorber and the condenser. Results show that it is possible to produce power and cooling at the same time without affecting the cooling output. Results also show that for all adsorbents used as the heat source temperature increases, the cooling effect and power generated increase. As for increasing the cold bed temperature, this will decrease the cooling effect and power output except for SAPO-34 which shows slightly increasing trend of cooling and power output. As the condenser cooling temperature increases, the cooling effect and power output will decrease while for the chilled water temperature, the cooling load and power generated increased as the temperature increased. The maximum values of average specific power generation (SP), specific cooling power (SCP) and cycle efficiency are 73 W

  2. The development of the gas cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Kernforschungszentrum Karlsruhe

    1975-01-01

    A survey of the present technological state is given on the basis of the developments made so far. Some milestones of development - e.g. the German gas breeder memorandum, the Gas Breeder Reactor Association the results of the BR-2 radiation experiments and of GfK-KWU design and safety studies - are described. The problems connected with a large store of plutonium are also discussed. (UA/AK) [de

  3. Operator training facilities for CEGB advanced gas cooled reactors

    International Nuclear Information System (INIS)

    Green, J.F.; Birnie, S.

    1980-01-01

    The facilities provided at the Nuclear Power Training Centre of the CEGB for the training of operators fo the AGR are described. The simulator control desks are replicas of three AGR designs with, in addition, simulation of the Data Processing System for each station. Three modes of operation are envisaged: a.) Demonstration where the simulator is used by the tutor to illustrate lecture on plant behaviour. b.) Interaction where the student carries out normal procedures and experiences plant failure situations. c.) Investigation where engineering staff use the simulator for validation of modified operational procedures, ergonomic studies etc. (orig./HP)

  4. Dry fuel store for advanced gas cooled reactor fuels

    International Nuclear Information System (INIS)

    Grant, J.S.; Boocock, P.M.; Ealing, C.J.

    1992-01-01

    This paper summarizes the fuel storage requirements in Scotland and the selection of a Dry Fuel Store of the Modular Vault Dry Store (MVDS) design developed by GEC ALSTHOM Engineering Systems Limited (GECA). A similar design of store has been selected and has been constructed in the USA by Foster Wheeler Energy Corporation in collaboration with GECA

  5. Computational Analysis of Supercritical Carbon Dioxide Gas Turbine for Liquid Metal Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Wi S.; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

    2008-10-15

    Energy demands at a remote site are increased as the world energy requirement diversifies so that they should generate power on their own site. A Small Modular Reactor (SMR) becomes a viable option for these sites. Generally, the economic feasibility of a high power reactor is greater than that for SMR. As a result the supercritical fluid driven Brayton cycle is being considered for a power conversion system to increase economic competitiveness of SMR. The Brayton cycle efficiency is much higher than that for the Rankine cycle. Moreover, the components of the Brayton cycle are smaller than Rankine cycle's due to high heat capacity when a supercritical fluid is adopted. A lead (Pb) cooled SMR, BORIS, and a supercritical fluid driven Brayton cycle, MOBIS, are being developed at the Seoul National University (SNU). Dostal et al. have compared some advanced power cycles and proposed the use of a supercritical carbon dioxide (SCO{sub 2}) driven Brayton cycle. According to their suggestion SCO{sub 2} is adopted as a working fluid for MOBIS. The turbo machineries are most important components for the Brayton cycle. The turbo machineries of Brayton cycle consists of a turbine to convert kinetic energy of the fluid into mechanical energy of the shaft, and a compressor to recompress and recover the driving force of the working fluid. Therefore, turbine performance is one of the pivotal factors in increasing the cycle efficiency. In MOBIS a supercritical gas turbine is designed in the Gas Advanced Turbine Operation (GATO) and analyzed in the Turbine Integrated Numerical Analysis (TINA). A three-dimensional (3D) numerical analysis is employed for more detailed design to account for the partial flow which the one-dimensional (1D) analysis cannot consider.

  6. Computational Analysis of Supercritical Carbon Dioxide Gas Turbine for Liquid Metal Cooled Reactor

    International Nuclear Information System (INIS)

    Jeong, Wi S.; Suh, Kune Y.

    2008-01-01

    Energy demands at a remote site are increased as the world energy requirement diversifies so that they should generate power on their own site. A Small Modular Reactor (SMR) becomes a viable option for these sites. Generally, the economic feasibility of a high power reactor is greater than that for SMR. As a result the supercritical fluid driven Brayton cycle is being considered for a power conversion system to increase economic competitiveness of SMR. The Brayton cycle efficiency is much higher than that for the Rankine cycle. Moreover, the components of the Brayton cycle are smaller than Rankine cycle's due to high heat capacity when a supercritical fluid is adopted. A lead (Pb) cooled SMR, BORIS, and a supercritical fluid driven Brayton cycle, MOBIS, are being developed at the Seoul National University (SNU). Dostal et al. have compared some advanced power cycles and proposed the use of a supercritical carbon dioxide (SCO 2 ) driven Brayton cycle. According to their suggestion SCO 2 is adopted as a working fluid for MOBIS. The turbo machineries are most important components for the Brayton cycle. The turbo machineries of Brayton cycle consists of a turbine to convert kinetic energy of the fluid into mechanical energy of the shaft, and a compressor to recompress and recover the driving force of the working fluid. Therefore, turbine performance is one of the pivotal factors in increasing the cycle efficiency. In MOBIS a supercritical gas turbine is designed in the Gas Advanced Turbine Operation (GATO) and analyzed in the Turbine Integrated Numerical Analysis (TINA). A three-dimensional (3D) numerical analysis is employed for more detailed design to account for the partial flow which the one-dimensional (1D) analysis cannot consider

  7. HTGR [High Temperature Gas-Cooled Reactor] ingress analysis using MINET

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Yang, J.W.; Kroeger, P.G.; Mallen, A.N.; Aronson, A.L.

    1989-04-01

    Modeling of water/steam ingress into the primary (helium) cooling circuit of a High Temperature Gas-Cooled Reactor (HTGR) is described. This modeling was implemented in the MINET Code, which is a program for analyzing transients in intricate fluid flow and heat transfer networks. Results from the simulation of a water ingress event postulated for the Modular HTGR are discussed. 27 refs., 6 figs., 6 tabs

  8. Cool infalling gas and its interaction with the hot ISM of elliptical galaxies

    Science.gov (United States)

    Sparks, W. B.; Macchetto, F. D.

    1990-01-01

    The authors describe work leading to the suggestion that interaction between infalling cool gas and ambient hot, coronal plasma in elliptical galaxies is responsible for emission filaments, and might remove the need for large mass depositions in cooling flows. A test of the hypothesis is undertaken - the run of surface brightness with radius for the emission lines - and the prediction agrees well with the data.

  9. Alternative nuclear fuel cycles and gas-cooled breeders

    International Nuclear Information System (INIS)

    Pellaud, B.; Dahlberg, R.C.

    1979-01-01

    The authors assert that it is prudent to concurrently develop more than one breeder system and to develop advanced converter reactors along with breeders. They provide arguments to support these assertions. (Auth.)

  10. Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1992-01-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

  11. Improvement of the decay heat removal characteristics of the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Epiney, A. S.

    2010-09-01

    The majority of NPPs worldwide are currently light water reactors, using ordinary water as both coolant and moderator. (...) For the longer-term future, viz. beyond the year 2030, Research and Development is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Since the very beginning of the international cooperation on Generation IV, viz. the year 2000, the main research interest in Europe as regards the advanced fast-spectrum systems needed for achieving complete fuel cycle closure, has been for the Sodium-cooled Fast Reactor (SFR). However, the Gas-cooled Fast Reactor (GFR) is currently considered as the main back-up solution. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For depressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure conditions, need to be

  12. Conceptual design study for the enhanced gas cooled reactor (EGCR)

    International Nuclear Information System (INIS)

    Nakano, M.; Sadahiro, D.; Ozaki, H.; Bryant, S.D.; Cheyne, A.; Gilroy, J.E.; Hulme, G.; Lennox, T.A.; Sunderland, R.E.; Beaumont, H.M.; Kida, M.; Nomura, M.

    2001-01-01

    The preliminary concept of the carbon dioxide cooled fast reactor EGCR has been studied as a Generation IV system. EGCR with MOX fuel has a very good core performance, a breeding ratio over 1.2, a long operating cycle of 24 months, and a high burnup of 150 GWd/t. The plant system is based on the successful AGR experience but provides 3600 MWth. Enhanced passive safety features are provided and a debris tray included. Preliminary costing studies show that EGCR can be competitive to LWRs and can be constructed on a similar schedule. This EGCR concept also shows development potential. (author)

  13. Construction and performance tests of a secondary hydrogen gas cooling system

    International Nuclear Information System (INIS)

    Sanokawa, K.; Hishida, M.

    1980-01-01

    With the aim of a multi-purpose use of nuclear energy, such as direct steel-making, an experimental multi-purpose high-temperature gas-cooled reactor (VHTR) is now being developed by the Japan Atomic Energy Research Institute (JAERI). In order to simulate a heat exchanging system between the primary helium gas loop and the secondary reducing gas system of the VHTR, a hydrogen gas loop as a secondary cooling system of the existing helium gas loop was completed in 1977, and was successfully operated for over 2000 hours. The objectives of constructing the H 2 secondary loop were: (1) To get basic knowledge for designing, constructing and operating a high-temperature and high-pressure gas facility; (2) To perform the following tests: (a) hydrogen permeation at the He/H 2 heat exchanger (the surfaces of the heat exchanger tubes are coated by calorizing to reduce hydrogen permeation), (b) thermal performance tests of the He/H 2 heat exchanger and the H 2 /H 2 regenerative heat exchanger, (c) performance test of internal insulation, and (d) performance tests of the components such as a H 2 gas heater and gas purifiers. These tests were carried out at He gas temperature of approximately 1000 0 C, H 2 gas temperature of approximately 900 0 C and gas pressures of approximately 40 kg/cm 2 G, which are almost the same as the operating conditions of the VHTR

  14. Steam generator materials constraints in UK design gas-cooled reactors

    International Nuclear Information System (INIS)

    James, D.W.

    1988-01-01

    A widely reported problem with Magnox-type reactors was the oxidation of carbon steel components in gas circuits and steam generators. The effects of temperature, pressure, gas composition and steel composition on oxidation kinetics have been determined, thus allowing the probabilities of failure of critical components to be predicted for a given set of operating conditions. This risk analysis, coupled with regular inspection of reactor and boiler internals, has allowed continued operation of all U.K. Magnox plant. The Advanced Gas Cooled Reactor (AGR) is a direct development of the Magnox design. The first four AGRs commenced operation in 1976, at Hinkley Point 'B' and at Hunterston 'B'. All known materials problems with the steam generators have been diagnosed and solved by the development of appropriate operational strategies, together with minor plant modifications. Materials constraints no longer impose any restrictions to full load performance from the steam generators throughout the predicted life of the plant. Problems discussed in detail are: 1. oxidation of the 9 Cr - 1 Mo superheater. 2. Stress corrosion of the austenitic superheater. 3. Creep of the transition joints between the 9 Cr - 1 Mo and austenitic sections. With the 9 Cr - 1 Mo oxidation maximum temperature restriction virtually removed and creep constraints properly quantified, boiler operation in now favourably placed. Stress corrosion research has allowed the risk of tube failure to be related to time, temperature, stress and chemistry. As a result, the rigorous 'no wetting' policy has been relaxed for the normally high quality AGR feedwater, and the superheat margin has been reduced to 23 deg. C. This has increased the size of the operating window and reduced the number of expensive, and potentially harmful, plant trips. (author)

  15. Single-jet gas cooling of in-beam foils or specimens: Prediction of the convective heat-transfer coefficient

    Science.gov (United States)

    Steyn, Gideon; Vermeulen, Christiaan

    2018-05-01

    An experiment was designed to study the effect of the jet direction on convective heat-transfer coefficients in single-jet gas cooling of a small heated surface, such as typically induced by an accelerated ion beam on a thin foil or specimen. The hot spot was provided using a small electrically heated plate. Heat-transfer calculations were performed using simple empirical methods based on dimensional analysis as well as by means of an advanced computational fluid dynamics (CFD) code. The results provide an explanation for the observed turbulent cooling of a double-foil, Havar beam window with fast-flowing helium, located on a target station for radionuclide production with a 66 MeV proton beam at a cyclotron facility.

  16. A directly cooled grating substrate for ALS [Advanced Light Source] undulator beam lines

    International Nuclear Information System (INIS)

    DiGennaro, R.; Swain, T.

    1989-08-01

    Design analyses using finite element methods are presented for thermal distortion of water-cooled diffraction grating substrates for a potential application at the LBL Advanced Light Source, demonstrating that refinements in cooling channel configuration and heat flux distribution can significantly reduce optical surface distortion with high heat loads. Using an existing grating substrate design, sensitivity of tangential slope errors due to thermal distortion is evaluated for a variety of thermal boundary conditions, including coolant flow rate and heat transfer film coefficients, surface illumination area and heat distribution profile, and location of the convection cooling surfaces adjacent to the heated region. 1 ref., 5 figs., 2 tabs

  17. Advances in passive cooling design and performance analysis

    International Nuclear Information System (INIS)

    Woodcock, J.

    1994-01-01

    The Third International Conference on Containment Design and Operation continues the trend of rapidly extending the state of the art in containment methodology, joining other conferences, OECD-sponsored International Standard Problem exercises, and vendor licensing submittals. Methodology developed for use on plants with passive features is under increasing scrutiny for advanced designs, since the passive features are often the only deviation from existing operating base of the past 30 years of commercial nuclear power. This session, 'Containment Passive Safety Systems Design and Operation,' offers papers on a wide range of topics, with authors from six organizations from around the world, dealing with general passive containments, Westinghouse AP600, large (>1400 MWe) passive plants, and the AECL advanced CANDU reactor. This level and variety of participation underscores the high interest and accelerated methods development associated with advanced passive containment heat removal. The papers presented in this session demonstrate that significant contributions are being made to the advancement of technology necessary for building a new generation of safer, more economical nuclear plants. (author)

  18. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    International Nuclear Information System (INIS)

    Hassan, Yassin; Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-01-01

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  19. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  20. Cooling of Gas Turbines. 6; Computed Temperature Distribution Through Cross Section of Water-Cooled Turbine Blade

    Science.gov (United States)

    Livingood, John N. B.; Sams, Eldon W.

    1947-01-01

    A theoretical analysis of the cross-sectional temperature distribution of a water-cooled turbine blade was made using the relaxation method to solve the differential equation derived from the analysis. The analysis was applied to specific turbine blade and the studies icluded investigations of the accuracy of simple methods to determine the temperature distribution along the mean line of the rear part of the blade, of the possible effect of varying the perimetric distribution of the hot gas-to -metal heat transfer coefficient, and of the effect of changing the thermal conductivity of the blade metal for a constant cross sectional area blade with two quarter inch diameter coolant passages.

  1. Properties of super alloys for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Izaki, Takashi; Nakai, Yasuo; Shimizu, Shigeki; Murakami, Takashi

    1975-01-01

    The existing data on the properties at high temperature in helium gas of iron base super alloys. Incoloy-800, -802 and -807, nickel base super alloys, Hastelloy-X, Inconel-600, -617 and -625, and a casting alloy HK-40 were collectively evaluated from the viewpoint of the selection of material for HTGRs. These properties include corrosion resistance, strength and toughness, weldability, tube making, formability, radioactivation, etc. Creep strength was specially studied, taking into consideration the data on the creep characteristics in the actual helium gas atmosphere. The necessity of further long run creep data is suggested. Hastelloy-X has completely stable corrosion resistance at high temperature in helium gas. Incoloy 800 and 807 and Inconel 617 are not preferable in view of corrosion resistance. The creep strength of Inconel 617 extraporated to 1,000 deg C for 100,000 hours in air was the greatest rupture strength of 0.6 kg/mm 2 in all above alloys. However, its strength in helium gas began to fall during a relatively short time, so that its creep strength must be re-evaluated in the use for long time. The radioactivation and separation of oxide film in primary construction materials came into question, Inconel 617 and Incoloy 807 showed high induced radioactivity intensity. Generally speaking, in case of nickel base alloys such as Hastelloy-X, oxide film is difficult to break away. (Iwakiri, K.)

  2. Development of cooling and cleaning systems for enhanced gas ...

    African Journals Online (AJOL)

    DR OKE

    Energy from biomass based gasifier-engine integrated systems are ..... an almost linear electrical signal of 40 /lVrc. ..... The diesel and ROME can be used as injected fuel in dual fuel mode with producer gas induction and this feature does not.

  3. Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

    International Nuclear Information System (INIS)

    Han, Kyu Hyun; Chang, Soon Heung

    2005-01-01

    More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

  4. Simulation of the fuzzy-smith control system for the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Li Deheng; Xu Xiaolin; Zheng Jie; Guo Renjun; Zhang Guifen

    1997-01-01

    The Fuzzy-Smith pre-estimate controller to solve the control of the big delay system is developed, accompanied with the development of the mathematical model of the 10 MW high temperature gas cooled test reactor (HTR-10) and the design of its control system. The simulation results show the Fuzzy-Smith pre-estimate controller has the advantages of both fuzzy control and Smith pre-estimate controller; it has better compensation to the delay and better adaptability to the parameter change of the control object. So it is applicable to the design of the control system for the high temperature gas cooled reactor

  5. ETDR, The European Union's Experimental Gas-Cooled Fast Reactor Project

    International Nuclear Information System (INIS)

    Poette, Christian; Brun-Magaud, Valerie; Morin, Franck; Dor, Isabelle; Pignatel, Jean-Francois; Bertrand, Frederic; Stainsby, Richard; Pelloni, Sandro; Every, Denis; Da Cruz, Dirceu

    2008-01-01

    In the Gas-Cooled Fast Reactor (GFR) development plan, the Experimental Technology Demonstration Reactor (ETDR) is the first necessary step towards the electricity generating prototype GFR. It is a low power (∼50 MWth) Helium cooled fast reactor. The pre-conceptual design of the ETDR is shared between European partners through the GCFR Specifically Targeted Research Project (STREP) within the European Commission's 6. R and D Framework Program. After recalling the place of ETDR in the GFR development plan, the main reactor objectives, the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: - Sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding). - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the starting core. - Starting Core reactivity management studies model including experimental GFR sub-assemblies. - Neutron and radiation shielding calculations using a specific MCNP model. The model allows evaluation of the neutron doses for the vessel and internals and radiation doses for maintenance operations. - System design and safety considerations, with a reactor architecture largely influenced by the Decay Heat Removal strategy (DHR) for de-pressurized accidents. The design of the reactor raises a number of issues in terms of fuel, neutronics, thermal-hydraulics codes qualification as well as critical components (blowers, IHX, thermal barriers) qualification. An overview of the R and D development on codes and technology qualification program is presented. Finally, the status of international collaborations and their perspectives for the ETDR are mentioned. (authors)

  6. Reactor Physics Development for Advanced Gas-Cooled Reactors; Recherches en Physique des Reacteurs, pour des Reacteurs Perfectionnes Refroidis par un Gaz; Razrabotka metodov v oblasti reaktornoj fiziki dlya usovershenstvovannogo reaktora s gazovym okhlazhdeniem; Progresos de la Fisica de los Reactores de Tipo Avanzado Refrigerados por Gas

    Energy Technology Data Exchange (ETDEWEB)

    Moore, J. [United Kingdom Atomic Energy Authority (United Kingdom)

    1964-04-15

    The paper is a general one reviewing the detailed experimental and theoretical work that has been done during design, development and commissioning of the Windscale AGR. and in general support of development of Advanced Gas-Cooled Reactors for civil power production. The main purpose of the paper is to describe die considerable amount of work that has been necessary to develop adequate theoretical methods for calculating (a) flux distribution and reactivity balance in a complex reactor core, (b) power distribution in complex fuel geometries, and (c) the effect of irradiation on fuel cycles and power distribution. As an introduction reference is made to experimental information and theoretical methods available from work on uranium Magnox systems and the experimental information obtained from the British Industries Collaborative Experimental Programme (BICEP) both of which provided the starting point for development of theoretical methods applicable to AGRs. The approach to critical facility APEX and the zero energy reactor HERO have been used with regular lattice arrangements and with combinations of perturbers such as control rods in them to determine lattice parameters for AGR and to check the theoretical methods developed for dealing with heterogeneous reactor cores. The theoretical methods that have been developed and used to date are known as hetrecontrol and FTD2. Experiments were designed to check detailed features of these methods and measurements on a number of different sized ''reactor'' cores in APEX and HERO were analysed to determine a self-consistent set of lattice constants that fitted the experimental results. These purely empirical constants were the used in hetrecontrol and FTD2 to successfully plan commissioning and choice of the loading pattern for the Windscale AGR. Reference is made to experimental techniques that have been tested and those developed for the particular problems encountered. The methods examined for measurement of reactivity

  7. Implementation of gas district cooling and cogeneration systems in Malaysia; Mise en oeuvre de systemes de gas district cooling et de cogeneration en Malaisie

    Energy Technology Data Exchange (ETDEWEB)

    Haron, S. [Gas District Cooling, M, Sdn Bhd (Malaysia)

    2000-07-01

    With its energy demand in the early 1990's growing at a high rate due to the country's strong economic growth, Malaysia studied various options to improve the efficiency of its energy use. Since its natural gas reserves are almost four times that of its crude oil reserves, efforts were therefore centered on seeking ways to boost the use of natural gas to mitigate the growing domestic energy need. PETRONAS, the national oil company, subsequently studied and chose the District Cooling System using natural gas as the primary source of fuel. The Kuala Lumpur City Center development, which houses the PETRONAS Twin Towers, was subsequently chosen as the first project to use the Gas District Cooling (GDC) System. To acquire the technology and implement this project, PETRONAS created a new subsidiary, Gas District Cooling (Malaysia) Sendirian Berhad (GDC(M)). In the process of improving the plant's efficiency, GDC(M) discovered that the GDC system's efficiency and project economics would be significantly enhanced if its is coupled to a Cogeneration system. Having proven the success of the GDC/Cogeneration system, GDC(M) embarked on a campaign to aggressively promote and seek new opportunities to implement the system, both in Malaysia-and abroad. Apart from enhancing efficiency of energy use, and providing better project economics, the GDC/Cogeneration system also is environment friendly. Today, the GDC/Cogeneration systems is the system of choice for several important developments in Malaysia, which also includes the country's prestigious projects such as the Kuala Lumpur International Airport and the New Federal Government Administrative Center in Putrajaya. (author)

  8. Simulation Of The Secondary Cooling System Failed For One Line Mode Of RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto; Susyadi; Sembiring, Tagor M; Isnaeni, Darwis

    2003-01-01

    Recently, an assessment of 15 MW power reactor RSG-GAS operated using one line cooling mode is under carried out, in which is in the same manner as BA TAN policy. At the power above mentioned, requirement for the research as well as isotop production has been fulfilled. To obtain the transient condition of 1 line-cooling mode, the simulation using RELAP5.MOD3.2 code was carried out. The simulation parameters interesting known are the inlet of primary coolant temperature after failed the secondary cooling system. At the first, reactor is operated at 15 MW steady state condition using 1 line-cooling mode. Primary coolant flow rate of 430 kg/s and secondary of 550 kg/s respectively. After that the decreasing is occurred due to stop of secondary cooling pump. Therefore the primary cooling inlet temperature to the core increase cause scram reactor by inserted control rod. During the transient occur, the characteristic of primary cooling temperature pattern change were obtained. The simulation result shows that the temperature increase (ΔT) temperature to the reactor is 5,1 o C at the second of 85.5. Here is lower than ΔT for the two cooling mode of 10 o C. That temperature characteristic still tolerable against acceptable safety margin to the flow instability

  9. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1976-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100MWd/kg-heavy metal. The fuel is a sol-gel-derived 88 at.% uranium (approximately 9% 235 U) and 12 at.% plutonium oxide, and the cladding is 20% cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70MWd/kg. The fuel has been operated at linear power rates of 39 and 44kW/m, and peak outer cladding temperature of 565 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48kW/m (685 0 C). Helium gas sweeps through any portion of the three regions of the fuel rod, namely: fuel, blanket, and charcoal trap. The charcoal trap is operated at about 300 0 C. An on-line Ge(Li) detector is used to analyse release rates of several gamma-emitting noble gas isotopes. Analyses are performed primarily on sweep gas flowing through the entire fuel rod, and for sweeps over the top of the charcoal trap. Sweep gas samples are analyzed for stable noble gas isotopes. Results in the form of ratios of release rate over birth rate (R/B) and venting rate over birth rate (V/B) are derived. R/B rates range from 10 -4 % to 30% while V/B ranges from 10 -6 % to 30%. Flow conductance in the capsule was monitored by recording the flow rate and pressure drop across the fuel rod and inlet sweep line. The flow conductance has been falling with increasing burnup, currently restricting the flow to about 20ml (s.t.p.)/min at a pressure difference of about 1.5MPa. Venting rates of the gaseous fission products as a function of gas pressure in the range 6.9 to 1.4MPa have also been measured. Planned future experiments include the monitoring of tritium release, venting and cladding permeation rates, and its molecular form. First measurements have been made. A simulated leak experiment will determine the mixture of fission gases as a function of flow rate and the most

  10. Progress in development and design aspects of advanced water cooled reactors

    International Nuclear Information System (INIS)

    1992-12-01

    The objective of the Technical Committee Meeting (TCM) was to provide an international forum for technical specialists to review and discuss technology developments and design work for advanced water cooled reactors, safety approaches and features of current water cooled reactors and to identify, understand and describe advanced features for safety and operational improvements. The TCM was attended by 92 participants representing 18 countries and two international organizations and included 40 presentations by authors of 14 countries and one international organization. A separate abstract was prepared for each of these presentations. Refs, figs, tabs

  11. Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

  12. Freezing around a pipeline carrying cooled gas in flooded areas

    Energy Technology Data Exchange (ETDEWEB)

    Koval' kov, V P; Krivoshein, B L

    1978-12-01

    The USSR's NIPIESUneftegazstroi mathematically analyzed the problem of ice formation around a subcooled-gas pipeline submerged in water in cold regions and derived charts for determining heat-transfer coefficients and the rate of ice formation for various water and gas temperatures. Because the ice halo that forms around these pipelines necessitates additional anchoring of the line, NIPIESUneftegazstroi sought to quantify the weight required in order to minimize the cost and material needed. The differential heat-transfer equations given can be used to calculate heat-transfer coefficients and the specific heat flux from the water to the ice halo, as well as the radius of the ice halo. Values of the ice-halo radius are plotted graphically as parabolic function of time (to 15,000 h) for pipeline surface temperatures of 30.2, 27.5, 23, 18.5, and 14/sup 0/F. An equation indicates the limiting value of the temperature of the transported gas at which icing of an insulated pipeline will not occur.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-15

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

  14. Assessment of the effect of nitrogen gas on passive containment cooling system performance

    International Nuclear Information System (INIS)

    Ha, Huiun; Suh, Jungsoo

    2016-01-01

    As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR

  15. Assessment of the effect of nitrogen gas on passive containment cooling system performance

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Suh, Jungsoo [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR.

  16. Heating and cooling system for an on-board gas adsorbent storage vessel

    Science.gov (United States)

    Tamburello, David A.; Anton, Donald L.; Hardy, Bruce J.; Corgnale, Claudio

    2017-06-20

    In one aspect, a system for controlling the temperature within a gas adsorbent storage vessel of a vehicle may include an air conditioning system forming a continuous flow loop of heat exchange fluid that is cycled between a heated flow and a cooled flow. The system may also include at least one fluid by-pass line extending at least partially within the gas adsorbent storage vessel. The fluid by-pass line(s) may be configured to receive a by-pass flow including at least a portion of the heated flow or the cooled flow of the heat exchange fluid at one or more input locations and expel the by-pass flow back into the continuous flow loop at one or more output locations, wherein the by-pass flow is directed through the gas adsorbent storage vessel via the by-pass line(s) so as to adjust an internal temperature within the gas adsorbent storage vessel.

  17. The world trends of high temperature gas-cooled reactors and the mode of utilization

    International Nuclear Information System (INIS)

    Ishikawa, Hiroshi; Shimokawa, Jun-ichi

    1974-01-01

    After a long period of research and development, high temperature gas-cooled reactors are going to enter the practical stage. The combination of a HTGR with a closed cycle helium gas turbine is advantageous in thermal efficiency, reduction of environmental impact and economy. In recent years, the direct utilization of nuclear heat energy in industries has been attracting interest. The multi-purpose utilization of high temperature gas-cooled reactors is thus now the world trend. Reviewing the world developments in this field, the following matters are described: (1) development of HTGRs in the U.K., West Germany, France and the United States; (2) development of He gas turbine, etc. in West Germany; and (3) multi-purpose utilization of HTGRs in West Germany and Japan. (Mori, K.)

  18. A UKAEA review of gas-cooled reactors in the United Kingdom

    International Nuclear Information System (INIS)

    Heath, E.C.; Knowles, A.N.

    1983-01-01

    The commercial use of nuclear power for electrical generation commenced in the UK in the 1950s with the Calder Hall reactors. Based on this concept, eighteen commercial reactor units, with two further units outside the UK, were constructed and have been in operation for periods ranging from 10 to 19 years. The paper reviews this experience mainly from the aspects of safety and the achieved costs, which compare favourably with current figures for fossil fired generation. The further development of the gas-cooled system in the UK commenced with the construction of the Windscale AGR, which came into operation in 1962. This led to the ordering of 14 large commercial AGR units, 4 of which have been in service since 1976, 6 are at an advanced stage of construction and 4 are at an early stage of construction. The paper reviews the main safety features of the AGR and considers the costs, taking achieved costs for the units which are in service and a combination of historical costs and projected costs for the units under construction. Again a clear advantage over fossil fuelled stations is shown. The paper also includes a preliminary account of the use of the prototype AGR at Windscale for the series of experiments concerning plateout, over-temperature in the fuel and simulated fault transients in the core which were carried out earlier in 1981. (author)

  19. The design status of the United States Department of Energy modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Mills, Raymond R. Jr.

    1990-01-01

    The U.S. Department of Energy's Modular High Temperature Gas Cooled Reactor (MHTGR) is being designed using a systems engineering approach referred to as the integrated approach. The top level requirement for the plant is that it provides safe, reliable, economical energy. The safety requirements are established by the U.S. Licensing Authorities, principally the Nuclear Regulatory Commission. The reliability and economic requirements associated with the top level functions have been established in close coordination and cooperation with the electrical utilities and other potential users, and the nuclear supply industry. The integrated approach uses functional analysis to define the functions and sub-functions for the plant and to identify quantitatively how the various functions must be fulfilled. The top four functions associated with the MHTGR are: maintain safe plant operation; maintain plant protection; maintain control of radionuclide release; maintain emergency preparedness. In addition to meeting all U.S. Regulatory Requirements this advanced reactor concept is being designed to meet the following requirements: do not require sheltering or evacuating of anyone outside the plant boundary of 425 meters as a result of normal or abnormal plant operation; do not require operator action in order to accomplish the above sheltering and evacuation objectives and the design must be insensitive to operator errors; utilize inherent characteristics of materials to develop passive safety features; provide very long times for corrective actions following the initiation of an abnormal event before plant damage would be incurred

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

    Directory of Open Access Journals (Sweden)

    Shutang Zhu

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail

  2. Unexpected mobility of OH+ and OD+ molecular ions in cooled helium gas

    International Nuclear Information System (INIS)

    Isawa, R; Yamazoe, J; Tanuma, H; Ohtsuki, K

    2012-01-01

    Mobilities of OH + and OD + ions in cooled helium gas have been measured at gas temperature of 4.3 K. Measured mobilities of both ions as a function of an effective temperature T eff show a minimum around 80 K, and they are approaching to the polarization limits at very low T eff . These findings will be related to the extremely strong anisotropy of the interaction potential between the molecular ion and helium atom.

  3. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements.

  4. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    International Nuclear Information System (INIS)

    1985-01-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

  5. Thermoregulated Nitric Cryosystem for Cooling Gas-Filled Detectors of Ionizing Radiation

    Directory of Open Access Journals (Sweden)

    Zharkov I.P.

    2015-09-01

    Full Text Available Cryosystem for cooling and filling of gas-filled detectors of ionizing radiation with compressed inert gas on the basis of wide-nitrogen cryostat, which provides detetector temperature control in a range of 173 — 293 K and its stabilization with accuracy of ± 1°. The work was carried out within the Ukraine — NATO Program of Collaboration, Grant SfP #984655.

  6. Auxiliary heat exchanger for a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Ecker, H.; Gasch, K.; Lischer, R.; Spilker, H.

    1978-01-01

    The proposal concerns the design configuration of the individual components of a heat exchanger with circular cross-section, being placed within a lined pod of the concrete shell of the pressure vessel. The heat exchanger has got a vertical cooler installed below the circulator. The components are arranged in such manner that the access to the pipe lines for in-service inspections is assured. Uniform velocity distribution of the gas streaming into the cooler from below is to be achieved. (GL) 891 GL/GL 892 MKO [de

  7. Simulation for temperature changing investigation at RSG-GAS cooling system

    International Nuclear Information System (INIS)

    Utaja

    2002-01-01

    The RSG-GAS cooling system considers of primary and secondary system, is used for heat rejection from reactor core to the atmosphere. For temperature changing investigation cause by atmospherics condition changing or coolant flow rate changing, is more safe done by simulation. This paper describes the simulation for determine the RSG-GAS coolant temperature changing base on heat exchange and cooling tower characteristic. The simulation is done by computer programme running under WINDOWS 95 or higher. The temperature changing is based on heat transfer process on heat exchanger and cooling tower. The simulation will show the water tank temperature changing caused by the temperature and humidity of the atmosphere or by coolant flow rate changing. For example the humidity changing from 60% to 80% atmospherics temperature 30 oC and 32400 k Watt power will change the tank temperature from 37,97 oC to 40,03 oC

  8. Advanced On Board Inert Gas Generation System (OBBIGS), Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Valcor Engineering Corporation proposes to develop an advanced On Board Inert Gas Generation System, OBIGGS, for aircraft fuel tank inerting to prevent hazardous...

  9. Condensation heat transfer with noncondensable gas for passive containment cooling of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leonardi, Tauna [Schlumberger, 14910 Airline Rd., Rosharon, TX 77583 (United States)]. E-mail: Tleonardi@slb.com; Ishii, Mamoru [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)]. E-mail: Ishii@ecn.purdue.edu

    2006-09-15

    Noncondensable gases that come from the containment and the interaction of cladding and steam during a severe accident deteriorate a passive containment cooling system's performance by degrading the heat transfer capabilities of the condensers in passive containment cooling systems. This work contributes to the area of modeling condensation heat transfer with noncondensable gases in integral facilities. Previously existing correlations and models are for the through-flow of the mixture of steam and the noncondensable gases and this may not be applicable to passive containment cooling systems where there is no clear passage for the steam to escape. This work presents a condensation heat transfer model for the downward cocurrent flow of a steam/air mixture through a condenser tube, taking into account the atypical characteristics of the passive containment cooling system. An empirical model is developed that depends on the inlet conditions, including the mixture Reynolds number and noncondensable gas concentration.

  10. Design activity of IHI on the experimental multipurpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    1978-01-01

    With conspicuous interest and attention paid by iron and steel manufacturing industries, the development of the multipurpose high temperature gas-cooled reactor, namely the process heat reactor has been energetically discussed in Japan. The experimental multipurpose high temperature gas-cooled reactor, planned by JAERI (the Japan Atomic Energy Research Institute), is now at the end of the adjustment design stage and about to enter the system synthesizing design stage. The design of the JAERI reactor as a pilot plant for process heat reactors that make possible the direct use of the heat, produced in the reactor, for other industrial uses was started in 1969, and has undergone several revisions up to now. The criticality of the JAERI reactor is expected to be realized before 1985 according to the presently published program. IHI has engaged in the developing work of HTGR (high temperature gas-cooled reactor) including VHTR (very high temperature gas-cooled reactor) for over seven years, producing several achievements. IHI has also participated in the JAERI project since 1973 with some other companies concerned in this field. The design activity of IHI in the development of the JAERI reactor is briefly presented in this paper. (auth.)

  11. Design of project management system for 10 MW high temperature gas-cooled test reactor

    International Nuclear Information System (INIS)

    Zhu Yan; Xu Yuanhui

    1998-01-01

    A framework of project management information system (MIS) for 10 MW high temperature gas-cooled test reactor is introduced. Based on it, the design of nuclear project management information system and project monitoring system (PMS) are given. Additionally, a new method of developing MIS and Decision Support System (DSS) has been tried

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

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-09-01

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

  13. Closed-cycle gas flow system for cooling a HTc dc-SQUID magnetometer

    NARCIS (Netherlands)

    Bosch, van den P.J.; Holland, H.J.; Brake, ter H.J.M.; Rogalla, H.

    1994-01-01

    A closed-cycle gas flow system for cooling a high-crit. temp. d.c.-superconducting quantum interference device (SQUID) magnetometer by means of a cryocooler has been designed, constructed and tested. The magnetometer is aimed to measure heart signals with a sensitivity of 0.1 pT/Hz1/2. The required

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

    International Nuclear Information System (INIS)

    1978-09-01

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

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

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-01-01

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

  16. Requirements for electricity producing gas-cooled reactors in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Schwarz, D.K.J.

    1989-01-01

    The paper describes requirements to a high-temperature gas-cooled reactor from the view-point of a utility in the Federal Republic of Germany. The requirements presented in the paper address different areas including plant size, availability, safety and economics. (author)

  17. Research and development for high temperature gas cooled reactor in Japan

    International Nuclear Information System (INIS)

    Taketani, K.

    1978-01-01

    The paper describes the current status of High Temperature Gas Cooled Reactor research and development work in Japan, with emphasis on the Experimental Very High Temperature Reactor (Exp. VHTR) to be built by Japan Atomic Energy Research Institute (JAERI) before the end of 1985. The necessity of construction of Exp. VHTR was explained from the points of Japanese energy problems and resources

  18. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2004-01-01

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  19. The modular high-temperature gas-cooled reactor (MHTGR)

    International Nuclear Information System (INIS)

    Neylan, A.J.

    1986-10-01

    The MHTGR is an advanced reactor concept being developed in the USA under a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes basic HTGR features of ceramic fuel, helium coolant and a graphite moderator. However the specific size and configuration are selected to utilize the inherently safe characteristics associated with these standard features coupled with passive safety systems to provide a significantly higher margin of safety and investment protection than current generation reactors. Evacuation or sheltering of the public is not required. The major components of the nuclear steam supply, with special emphasis on the core, are described. Safety assessments of the concept are discussed

  20. Ceramics technology for advanced industrial gas turbines

    International Nuclear Information System (INIS)

    Anson, D.; Sheppard, W.J.; DeCorso, M.; Parks, W.J. Jr.

    1991-01-01

    Recent developments in the fabrication of high strength ceramic materials and in their application to automotive and aerospace gas turbine engines may lead also to significant improvements in the performance of industrial gas turbines. This paper presents a brief review of the improvements projected in a study initiated by the U.S. Department of Energy. The future costs of power generated by small gas turbines (up to 25 MW) are predicted, as well as the potential for fuel savings. Gas turbines in this size range are used extensively for gas compression and for cogeneration, as well as in a variety of more diverse applications. This paper includes results of analyses of the ways in which changes in gas turbine cost and performance are likely to affect market penetration. These results lead to predictions of future savings in U.S. fuel consumption in the industrial sector that would result. The paper also presents a brief overview of the scope of a suggested R and D program, with an appropriate schedule, which would provide a technical basis for achieving the projected results. Important parts of this program would cover ceramic design and fabrication technology, engine development and demonstration, and combustion technology

  1. Experimental investigation of heat transfer and flow using V and broken V ribs within gas turbine blade cooling passage

    Science.gov (United States)

    Kumar, Sourabh; Amano, R. S.

    2015-05-01

    Gas turbines are extensively used for aircraft propulsion, land-based power generation, and various industrial applications. With an increase in turbine rotor inlet temperatures, developments in innovative gas turbine cooling technology enhance the efficiency and power output; these advancements of turbine cooling have allowed engine designs to exceed normal material temperature limits. For internal cooling design, techniques for heat extraction from the surfaces exposed to hot stream of gas are based on an increase in the heat transfer areas and on the promotion of turbulence of the cooling flow. In this study, an improvement in performance is obtained by casting repeated continuous V- and broken V-shaped ribs on one side of the two pass square channels into the core of the blade. A detailed experimental investigation is done for two pass square channels with a 180° turn. Detailed heat transfer distribution occurring in the ribbed passage is reported for a steady state experiment. Four different combinations of 60° V- and broken 60° V-ribs in a channel are considered. A series of thermocouples are used to obtain the temperature on the channel surface and local heat transfer coefficients are obtained for Reynolds numbers 16,000, 56,000 and 85,000 within the turbulent flow regime. Area averaged data are calculated in order to compare the overall performance of the tested ribbed surface and to evaluate the degree of heat transfer enhancement induced by the rib. Flow within the channels is characterized by heat transfer enhancing ribs, bends, rotation and buoyancy effects. A series of experimental measurements is performed to predict the overall performance of the channel. This paper presents an attempt to collect information about the Nusselt number, the pressure drop and the overall performance of the eight different ribbed ducts at the specified Reynolds number. The main contribution of this study is to evaluate the best combination of rib arrangements

  2. Preliminary study on helium turbomachine for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Chen Yihua; Wang Jie; Zhang Zuoyi

    2003-01-01

    In the high temperature gas-cooled reactor (HTGR), gas turbine cycle is a new concept in the field of nuclear power. It combines two technologies of HTGR and gas turbine cycle, which represent the state-of-the-art technologies of nuclear power and fossil fuel generation respectively. This approach is expected to improve safety and economy of nuclear power plant significantly. So it is a potential scheme with competitiveness. The heat-recuperated cycle is the main stream of gas turbine cycle. In this cycle, the work medium is helium, which is very different from the air, so that the design features of the helium turbomachine and combustion gas turbomachine are different. The paper shows the basic design consideration for the heat-recuperated cycle as well as helium turbomachine and highlights its main design features compared with combustion gas turbomachine

  3. A Techno-Economic Assessment of Hybrid Cooling Systems for Coal- and Natural-Gas-Fired Power Plants with and without Carbon Capture and Storage.

    Science.gov (United States)

    Zhai, Haibo; Rubin, Edward S

    2016-04-05

    Advanced cooling systems can be deployed to enhance the resilience of thermoelectric power generation systems. This study developed and applied a new power plant modeling option for a hybrid cooling system at coal- or natural-gas-fired power plants with and without amine-based carbon capture and storage (CCS) systems. The results of the plant-level analyses show that the performance and cost of hybrid cooling systems are affected by a range of environmental, technical, and economic parameters. In general, when hot periods last the entire summer, the wet unit of a hybrid cooling system needs to share about 30% of the total plant cooling load in order to minimize the overall system cost. CCS deployment can lead to a significant increase in the water use of hybrid cooling systems, depending on the level of CO2 capture. Compared to wet cooling systems, widespread applications of hybrid cooling systems can substantially reduce water use in the electric power sector with only a moderate increase in the plant-level cost of electricity generation.

  4. Status and prospects for gas cooled reactor fuels. Proceedings of two IAEA meetings held in June 2004 and June 2005

    International Nuclear Information System (INIS)

    2009-04-01

    Recently, efforts to develop high temperature gas cooled reactors with an aim to building futuristic nuclear energy systems with advanced nuclear fuel cycles in the context of the Generation IV International Forum have increased significantly. In addition, several development projects are ongoing, focusing on the burning of weapons grade plutonium, including civil plutonium and other transuranic elements using the 'deep-burn concept', or 'inert matrix fuels', especially in the form of coated particles in gas cooled reactor systems. There is also considerable global interest in developing 'nuclear hydrogen' energy systems using high temperature gas cooled reactors. Apart from these developments, the value of preserving the large technology base developed in Germany, the United Kingdom and the United States of America, as well as information developed in other countries, has also been a subject of interest to the IAEA. At the second annual meeting of the 'technical working group on nuclear fuel cycles options and spent fuel management' (TWG-NFCO), held in Vienna from 28-30 May 2003, it was recommended to hold a technical meeting on Current Status and Future Prospects of Gas Cooled Reactor Fuels. The meeting should cover the technological progress that has been made in the last three years and plan future fabrication and qualification facilities for GCR/HTR fuel. TWG-NFCO considered it timely that this progress should be presented and discussed in the interested community. Recognizing the numerous activities being pursued in many Member States, the IAEA convened the technical meeting on this topic in June 2004 in Vienna. Consequently, an update meeting was held in June 2005, which was hosted by the Kharkov Institute of Physics and Technology of Ukraine to review and integrate the latest developments. This publication combines the results of the technical meeting of June 2004 and the meeting of June 2005. The proceedings presented here contain 25 in depth papers on the

  5. The Advancement of Cool Roof Standards in China from 2010 to 2015

    Energy Technology Data Exchange (ETDEWEB)

    Ge, Jing [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Levinson, Ronnen M. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2016-11-01

    Since the initiation of the U.S.-China Clean Energy Research Center-Building Energy Efficiency (CERC-BEE) cool roof research collaboration between the Lawrence Berkeley National Laboratory Heat Island Group and Chinese institutions in 2010, new cool surface credits (insulation trade- offs) have been adopted in Chinese building energy efficiency standards, industry standards, and green building standards. JGJ 75-2012: Design Standard for Energy Efficiency of Residential Buildings in Hot Summer and Warm Winter Zone became the first national level standard to provide cool surface credits. GB/T 50378-2014: Assessment Standard for Green Building is the first national level green building standard that offers points for heat island mitigation. JGJ/T 359-2015: Technical Specification for Application of Architectural Reflective Thermal Insulation Coating is the first industry standard that offers cool coating credits for both public and residential buildings in all hot-summer climates (Hot Summer/Cold Winter, Hot Summer/Warm Winter). As of December 2015, eight provinces or municipalities in hot-summer regions have credited cool surfaces credits in their residential and/or public building design standards; five other provinces or municipalities in hot-summer regions recommend, but do not credit, the use of cool surfaces in their building design standards. Cool surfaces could be further advanced in China by including cool roof credits for residential and public building energy efficiency standards in all hot-summer regions; developing a standardized process for natural exposure and aged-property rating of cool roofing products; and adapting the U.S.-developed laboratory aging process for roofing materials to replicate solar reflectance changes induced by natural exposure in China.

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

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

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

  7. FY 1999 report on survey. Introducing gas cooling systems in the Sultanate of Oman

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-03-01

    This study is conducted with CDM (clean development mechanism), etc., taken into consideration. Gas-aided cooling systems are introduced into the Royal Hospital, and other buildings and areas, replacing the existing motor-driven cooling systems. Their introduction into the Royal Hospital will reduce energy consumption at a rate of 1,855 toe/year and greenhouse gas at a rate of 5,129 tCO2/year. The effect to result from diffusion of such systems into the Royal Hospital and into large buildings, which are located near the existing natural gas pipelines at the center or in the vicinity of Muscat City and produce heavy cooling loads, is assessed. The calculation result indicates that energy conservation of 13,049 toe/year and greenhouse gas reduction of 31,636 tCO2/year will be available at universities, hotels, government office sectors, etc. As for the investment recovery period, it will be 4.7-5.3 years in case the gas decompression station construction cost is not included and 6.2-6.8 years in case the cost is included. The new system will cost 7% less than the existing system. In terms of electricity bill, the new system will be charged 30% less than the existing system. The Omani Government is critical of Japan's plan of establishing an energy supply corporation at the site at issue (even if financed by Japan). (NEDO)

  8. Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Kim, D. W.; Ryu, W. S.; Han, C. H.; Yoon, J. H.; Chang, J.

    2005-01-01

    Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

  9. The AGT 101 advanced automotive gas turbine

    Science.gov (United States)

    Rackley, R. A.; Kidwell, J. R.

    1982-01-01

    A development program is described whose goal is the accumulation of the technology base needed by the U.S. automotive industry for the production of automotive gas turbine powertrains. Such gas turbine designs must exhibit reduced fuel consumption, a multi-fuel capability, and low exhaust emissions. The AGT101 powertrain described is a 74.6 kW, regenerated single-shaft gas turbine, operating at a maximum inlet temperature of 1644 K and coupled to a split differential gearbox and automatic overdrive transmission. The engine's single stage centrifugal compressor and single stage radial inflow turbine are mounted on a common shaft, and will operate at a maximum rotor speed of 100,000 rpm. All high temperature components, including the turbine rotor, are ceramic.

  10. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  11. Transient thermal-hydraulic simulations of direct cycle gas cooled reactors

    International Nuclear Information System (INIS)

    Tauveron, Nicolas; Saez, Manuel; Marchand, Muriel; Chataing, Thierry; Geffraye, Genevieve; Bassi, Christophe

    2005-01-01

    This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to 'active' heat extraction systems

  12. Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE

    International Nuclear Information System (INIS)

    Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.

    2003-01-01

    This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation

  13. Construction and performance testing of a secondary cooling system with hydrogen gas (I)

    International Nuclear Information System (INIS)

    Hishida, M.; Nekoya, S.; Takizuka, T.; Emori, K.; Ogawa, M.; Ouchi, M.; Okamoto, Y.; Sanokawa, K.; Nakano, T.; Hagiwara, T.

    1979-08-01

    An experimental multi-purpose High-Temperature Gas Cooled Reactor (VHTR) which is supposed to be used for a direct steel-making is now being developed in JAeRI. In order to simulate the heat exchanging system between the primary helium gas and the secondary reducing gas system of VHTR, a hydrogen gas loop was constructed as a secondary cooling system of the helium gas loop. The maximum temperature and the maximum pressure of the hydrogen gas are 900 degrees C and 42 kg/cm 2 x G respectively. The construction of the hydrogen gas loop was completed in January, 1977, and was successfully operated for 1.000 h. Various performance tests, such as the hydrogen permeation test of a He/H2 heat exchanger and the thermal performance test of heat exchangers, were made. Especially, it was proved that hydrogen permeation rate through the heat exchanger was reduced to 1/30 to approximately 1/50 by a method of calorized coating, and the coating was stable during 1.000 h's operation. It was also stable against the temperature changes. This report describes the outline of the facility and performance of the components. (orig.) [de

  14. IAEA activities in technology development for advanced water-cooled nuclear power plants

    International Nuclear Information System (INIS)

    Juhn, Poong Eil; Kupitz, Juergen; Cleveland, John; Lyon, Robert; Park, Je Won

    2003-01-01

    As part of its Nuclear Power Programme, the IAEA conducts activities that support international information exchange, co-operative research and technology assessments and advancements with the goal of improving the reliability, safety and economics of advanced water-cooled nuclear power plants. These activities are conducted based on the advice, and with the support, of the IAEA Department of Nuclear Energy's Technical Working Groups on Advanced Technologies for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs). Assessments of projected electricity generation costs for new nuclear plants have shown that design organizations are challenged to develop advanced designs with lower capital costs and short construction times, and sizes, including not only large evolutionary plants but also small and medium size plants, appropriate to grid capacity and owner financial investment capability. To achieve competitive costs, both proven means and new approaches should be implemented. The IAEA conducts activities in technology development that support achievement of improved economics of water-cooled nuclear power plants (NPPs). These include fostering information sharing and cooperative research in thermo-hydraulics code validation; examination of natural circulation phenomena, modelling and the reliability of passive systems that utilize natural circulation; establishment of a thermo-physical properties data base; improved inspection and diagnostic techniques for pressure tubes of HWRs; and collection and balanced reporting from recent construction and commissioning experiences with evolutionary water-cooled NPPs. The IAEA also periodically publishes Status Reports on global development of advanced designs. (author)

  15. Flow-induced and acoustically induced vibration experience in operating gas-cooled reactors

    International Nuclear Information System (INIS)

    Halvers, L.J.

    1977-03-01

    An overview has been presented of flow-induced and acoustically induced vibration failures that occurred in the past in gas-cooled graphite-moderated reactors, and the importance of this experience for the Gas-Cooled Fast-Breeder Reactor (GCFR) project has been assessed. Until now only failures in CO 2 -cooled reactors have been found. No problems with helium-cooled reactors have been encountered so far. It is shown that most of the failures occurred because flow-induced and acoustically induced dynamic loads were underestimated, while at the same time not enough was known about the influence of environmental parameters on material behavior. All problems encountered were solved. The comparison of the influence of the gas properties on acoustically induced and flow-induced vibration phenomena shows that the interaction between reactor design and the thermodynamic properties of the primary coolant precludes a general preference for either carbon dioxide or helium. The acoustic characteristics of CO 2 and He systems are different, but the difference in dynamic loadings due to the use of one rather than the other remains difficult to predict. A slight preference for helium seems, however, to be justified

  16. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs

  17. Hot gas path component having cast-in features for near wall cooling

    Science.gov (United States)

    Miranda, Carlos Miguel; Kottilingam, Srikanth Chandrudu; Lacy, Benjamin Paul

    2018-04-10

    A hot gas path component includes a substrate having an outer surface and an inner surface. The inner surface of the substrate defines at least one interior space. At least a portion of the outer surface of the substrate includes a recess formed therein. The recess includes a bottom surface and a groove extending at least partially along the bottom surface of the recess. A cover is disposed within the recess and covers at least a portion of the groove. The groove is configured to channel a cooling fluid therethrough to cool the cover.

  18. A combined heating cooling stage for cluster thermalization in the gas phase

    International Nuclear Information System (INIS)

    Ievlev, D.N.; Kuester, A.; Enders, A.; Malinowski, N.; Schaber, H.; Kern, K.

    2003-01-01

    We report on the design and performance of a combined heating/cooling stage for the thermalization of clusters in a gas phase time-of-flight mass spectrometer. With this setup the cluster temperature can sensitively be adjusted within the range from 100 up to 800 K and higher. The unique combination of a heating stage with a subsequent cooling stage allows us to perform thermodynamic investigations on clusters at very high temperatures without quality losses in the spectra due to delayed fragmentation in the drift tube of the mass spectrometer. The performance of the setup is demonstrated by the example of (C 60 ) n clusters

  19. Advanced Gas Storage Concepts: Technologies for the Future

    Energy Technology Data Exchange (ETDEWEB)

    Freeway, Katy (PB-KBB Inc.); Rogers, R.E. (Mississippi State University); DeVries, Kerry L.; Nieland, Joel D.; Ratigan, Joe L.; Mellegard, Kirby D. (RESPEC)

    2000-02-01

    This full text product includes: 1) A final technical report titled Advanced Underground Gas Storage Concepts, Refrigerated-Mined Cavern Storage and presentations from two technology transfer workshops held in 1998 in Houston, Texas, and Pittsburgh, Pennsylvania (both on the topic of Chilled Gas Storage in Mined Caverns); 2) A final technical report titled Natural Gas Hydrates Storage Project, Final Report 1 October 1997 - 31 May 1999; 3) A final technical report titled Natural Gas Hydrates Storage Project Phase II: Conceptual Design and Economic Study, Final Report 9 June - 10 October 1999; 4) A final technical report titled Commerical Potential of Natural Gas Storage in Lined Rock Caverns (LRC) and presentations from a DOE-sponsored workshop on Alternative Gas Storage Technologies, held Feb 17, 2000 in Pittsburgh, PA; and 5) Phase I and Phase II topical reports titled Feasibility Study for Lowering the Minimum Gas Pressure in Solution-Mined Caverns Based on Geomechanical Analyses of Creep-Induced Damage and Healing.

  20. Study on external reactor vessel cooling capacity for advanced large size PWR

    International Nuclear Information System (INIS)

    Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

    2014-01-01

    External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

  1. Internally gas-cooled radiofrequency applicators as an alternative to conventional radiofrequency and microwave ablation devices: An in vivo comparison

    Energy Technology Data Exchange (ETDEWEB)

    Rempp, Hansjörg, E-mail: Hansjoerg.rempp@med.uni-tuebingen.de [Eberhard Karls University of Tübingen, Tübingen University Hospital, Department of Diagnostic and Interventional Radiology, Hoppe-Seyler-Straße 3, Tübingen, 72076 (Germany); Voigtländer, Matthias [ERBE Elektromedizin GmbH, Waldhörnlestraße 17, 72072 Tübingen (Germany); Schenk, Martin [Eberhard Karls University of Tuebingen, Tübingen University Hospital, Department of General, Visceral and Transplant Surgery, Hoppe-Seyler-Straße 3, 72076 Tübingen (Germany); Enderle, Markus D. [ERBE Elektromedizin GmbH, Waldhörnlestraße 17, 72072 Tübingen (Germany); Scharpf, Marcus [Eberhard Karls University of Tuebingen, Insitute of Pathology, Department on General Pathology and Pathological Anatomy, Liebermeisterstraße 8, 72076 Tübingen (Germany); Greiner, Tim O. [Eberhard Karls University of Tuebingen, Tübingen University Hospital, Department of General, Visceral and Transplant Surgery, Hoppe-Seyler-Straße 3, 72076 Tübingen (Germany); Neugebauer, Alexander [ERBE Elektromedizin GmbH, Waldhörnlestraße 17, 72072 Tübingen (Germany); and others

    2013-08-15

    Purpose: To test the efficacy of internally CO{sub 2}-cooled radiofrequency (RF) ablation in vivo and to compare its effectiveness to a standard water-cooled RF probe and to a gas-cooled microwave (MW) device. Method and materials: 49 ablations were performed on 15 pigs under general anesthesia using 15G monopolar CO{sub 2}-cooled RF applicators, 17G monopolar water-cooled RF applicators and 15G internally CO{sub 2}-cooled microwave devices. The power of the MW device was 45 W, the current of the gas-cooled RF device was 1200–1600 mA. At the water-cooled RF probe, maximum power of 200 W was set. Ablation time was 15 min. The short and long axes of the ablation zone were measured. Histological analyses and NADH-staining were performed. The diameters and the ablation volumes were compared using an analysis of variance. Results: No spots of untreated tissue were observed close to the cooled needle track in any of the ablation zones. The largest short axis diameter was 3.4 ± 0.5 cm achieved with the gas-cooled monopolar applicator. With the water-cooled applicators, short axis diameter was significantly smaller, reaching 2.5 ± 0.4 cm. Gas-cooled MW probes achieved 2.9 ± 1.0 cm. The largest ablation volume was 31.5 ± 12 ml (gas-cooled RF), and the smallest was 12.7 ± 4 ml (water-cooled RF). Short/long axis ratio was largest for gas-cooled RF probes with 0.73 ± 0.08 versus 0.64 ± 0.04 for the water-cooled probes and 0.49 ± 0.25 for the microwave applicator. Conclusion: Gas-cooled RF applicators may have a higher potential for effective destruction of liver lesions than comparable water-cooled RF systems, and may be an alternative to standard RF and MW ablation devices.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

  4. A study of the cool gas in the Large Magellanic Cloud. I. Properties of the cool atomic phase - a third H i absorption survey

    Science.gov (United States)

    Marx-Zimmer, M.; Herbstmeier, U.; Dickey, J. M.; Zimmer, F.; Staveley-Smith, L.; Mebold, U.

    2000-02-01

    The cool atomic interstellar medium of the Large Magellanic Cloud (LMC) seems to be quite different from that in the Milky Way. In a series of three papers we study the properties of the cool atomic hydrogen in the LMC (Paper I), its relation to molecular clouds using SEST-CO-observations (Paper II) and the cooling mechanism of the atomic gas based on ISO-[\\CII]-investigations (Paper III). In this paper we present the results of a third 21 cm absorption line survey toward the LMC carried out with the Australia Telescope Compact Array (ATCA). 20 compact continuum sources, which are mainly in the direction of the supergiant shell LMC 4, toward the surroundings of 30 Doradus and toward the eastern steep \\HI\\ boundary, have been chosen from the 1.4 GHz snapshot continuum survey of Marx et al. We have identified 20 absorption features toward nine of the 20 sources. The properties of the cool \\HI\\ clouds are investigated and are compared for the different regions of the LMC taking the results of Dickey et al. (survey 2) into account. We find that the cool \\HI\\ gas in the LMC is either unusually abundant compared to the cool atomic phase of the Milky Way or the gas is clearly colder (\\Tc\\ ~ 30 K) than that in our Galaxy (\\Tc\\ ~ 60 K). The properties of atomic clouds toward 30 Doradus and LMC 4 suggest a higher cooling rate in these regions compared to other parts of the LMC, probably due to an enhanced pressure near the shock fronts of LMC 4 and 30 Doradus. The detected cool atomic gas toward the eastern steep \\HI\\ boundary might be the result of a high compression of gas at the leading edge. The Australia Telescope is funded by the Commonwealth of Australia for operation as a National Facility managed by CSIRO.

  5. Experimental Flow Performance Evaluation of novel miniaturized Advanced Piezoelectric Dual Cooling Jet

    International Nuclear Information System (INIS)

    De Bock, H P J; Whalen, B P; Chamarthy, P; Jackson, J L

    2012-01-01

    In recent years, electronics systems have significantly reduced in size at maintained or increased functionality. This trend has led to an increased demand for smaller and more capable thermal management. However, miniaturization of conventional fan and heat sink cooling systems introduce significant size, weight and efficiency challenges. In this study the flow performance of a novel alternative thin form-factor cooling solution, the advanced piezoelectric dual cooling jet(DCJ), is evaluated. A DCJ is a system where two piezoelectric actuators are excited to produce air flow. The total height of the device is about 1mm. The design of the experimental method for evaluating the equivalent fan-curve of the DCJ device is described in detail. Experimental results in comparison to conventional fan solutions are provided. The DCJ is expected to be a good candidate for thermal management in next generation thin profile consumer electronics.

  6. Flow structure and heat exchange analysis in internal cooling channel of gas turbine blade

    Science.gov (United States)

    Szwaba, Ryszard; Kaczynski, Piotr; Doerffer, Piotr; Telega, Janusz

    2016-08-01

    This paper presents the study of the flow structure and heat transfer, and also their correlations on the four walls of a radial cooling passage model of a gas turbine blade. The investigations focus on heat transfer and aerodynamic measurements in the channel, which is an accurate representation of the configuration used in aeroengines. Correlations for the heat transfer coefficient and the pressure drop used in the design of radial cooling passages are often developed from simplified models. It is important to note that real engine passages do not have perfect rectangular cross sections, but include corner fillet, ribs with fillet radii and special orientation. Therefore, this work provides detailed fluid flow and heat transfer data for a model of radial cooling geometry which possesses very realistic features.

  7. Effects of cooling timescale and non-ideaness of the gas in the shockwaves

    Directory of Open Access Journals (Sweden)

    Mohsen Nejad-Asghar

    2017-09-01

    Full Text Available According to the suddenly compression of the matters in some regions of the compressible fluids, the density and temperature suddenly increases, and shockwaves can be produced. The cooling of post-shock region and non-idealness of the equation of state, $p=(k_B/mu m_prho T (1+brho equivmathcal{K}rho T (1+eta R$, where $mu m_p$ is the relative density of the post-shock gas and $Requiv rho_2 / rho_1$ is the non-idealness parameter, may affect on the shocked gases. In this article, we study the effects of both cooling timescale and non-idealness of the shocked gases, on the relative density of the post-shock region. For simplicity, the shock is assumed planar and steady in which the deceleration is negligible and there is no any instabilities through the cooling layer. Conservation of mass, momentum, and energy across the shock front are given by the Rankine-Hugoniot conditions. The most important factor through the shock is the energy lost per unit mass during the shock process, $Q=frac{n_2 Lambda}{mu_2 m_p} t_{dur}$, where $Lambda (erg cm^{-3} s^{-1}$ is the cooling function at the post-shock region with density $n_2} and mean particle mass $mu_2 m_p$, and $t_{dur}$ is the duration time of the post-shock process. Accurate determination of the cooling timescale requires specifying the elemental abundance of the post-shock region, but a simple estimate can be obtained using $t_{cool}approx k_B T_2/(n_2Lambda$. Eliminating the $n_2 Lambda$, we approximately have $Q/c^2approx lambda T$, where $c equiv sqrt{K_1 T_1}$ is the pre-shock sound speed, $lambda  equiv t_{dur}/t_{cool}$ and $T equiv K_2 T_2/K_1 T_1$. We would be interested to consider the collision of two gas sheets with velocities $v_0$ in the rest frame of the laboratory. Defining the Mach number as $M_0 equiv v_0/c$, we obtain a third degree polynomial equation for $R$, with coefficients as functions of the three parameters $eta$, $lambda$, and $M_0$. We numerically solved this three

  8. ADVANCED STRIPPER GAS PRODUCED WATER REMEDIATION

    International Nuclear Information System (INIS)

    Ray W. Sheldon

    2001-01-01

    Natural gas and oil production from stripper wells also produces water contaminated with hydrocarbons, and in most locations, salts and trace elements. The hydrocarbons are not generally present in concentrations that allow the operator to economically recover these liquids. Produced liquids, (stripper gas water) which are predominantly water, present the operator with two options; purify the water to acceptable levels of contaminates, or pay for the disposal of the water. The project scope involves testing SynCoal as a sorbent to reduce the levels of contamination in stripper gas well produced water to a level that the water can be put to a productive use. Produced water is to be filtered with SynCoal, a processed sub-bituminous coal. It is expected that the surface area of and in the SynCoal would sorb the hydrocarbons and other contaminates and the effluent would be usable for agricultural purposes. Test plan anticipates using two well locations described as being disparate in the level and type of contaminates present. The loading capacity and the rate of loading for the sorbent should be quantified in field testing situations which include unregulated and widely varying liquid flow rates. This will require significant flexibility in the initial stages of the investigation. The scope of work outlined below serves as the guidelines for the testing of SynCoal carbon product as a sorbent to remove hydrocarbons and other contaminants from the produced waters of natural gas wells. A maximum ratio of 1 lb carbon to 100 lbs water treated is the initial basis for economic design. While the levels of contaminants directly impact this ratio, the ultimate economics will be dictated by the filter servicing requirements. This experimental program is intended to identify those treatment parameters that yield the best technological practice for a given set of operating conditions. The goal of this research is to determine appropriate guidelines for field trials by accurately

  9. Nuclear data uncertainty analysis for the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Pelloni, S.; Mikityuk, K.

    2012-01-01

    For the European 2400 MW Gas-cooled Fast Reactor (GoFastR), this paper summarizes a priori uncertainties, i.e. without any integral experiment assessment, of the main neutronic parameters which were obtained on the basis of the deterministic code system ERANOS (Edition 2.2-N). JEFF-3.1 cross-sections were used in conjunction with the newest ENDF/B-VII.0 based covariance library (COMMARA-2.0) resulting from a recent cooperation of the Brookhaven and Los Alamos National Laboratories within the Advanced Fuel Cycle Initiative. The basis for the analysis is the original GoFastR concept with carbide fuel pins and silicon-carbide ceramic cladding, which was developed and proposed in the first quarter of 2009 by the 'French alternative energies and Atomic Energy Commission', CEA. The main conclusions from the current study are that nuclear data uncertainties of neutronic parameters may still be too large for this Generation IV reactor, especially concerning the multiplication factor, despite the fact that the new covariance library is quite complete; These uncertainties, in relative terms, do not show the a priori expected increase with bum-up as a result of the minor actinide and fission product build-up. Indeed, they are found almost independent of the fuel depletion, since the uncertainty associated with 238 U inelastic scattering results largely dominating. This finding clearly supports the activities of Subgroup 33 of the Working Party on International Nuclear Data Evaluation Cooperation (WPEC), i.e. Methods and issues for the combined use of integral experiments and covariance data, attempting to reduce the present unbiased uncertainties on nuclear data through adjustments based on available experimental data. (authors)

  10. A comparative study on recycling spent fuels in gas-cooled fast reactors

    International Nuclear Information System (INIS)

    Choi, Hangbok; Baxter, Alan

    2010-01-01

    This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

  11. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  12. High temperature gas-cooled reactors - once-through fuel cycle

    International Nuclear Information System (INIS)

    1979-03-01

    The HTGR, because of a unique combination of design characteristics, is a resource-efficient and cost-effective reactor. In the HTGR, the low power density core, coated particle fuel design, and gas cooling combine to provide high neutron economy, fuel burnup and thermodynamic efficiency. The uranium resource requirements for the current MEU/Th cycle with annual refueling results in a 30-year net U 3 O 8 requirement of 4280 ST/GWe. The basic design of the HTGR refueling scheme, whereby only selected regions of the core need be accessible during each refueling, makes fuel utilization improvements through semi-annual refueling an acceptable alternative in terms of plant availability. This alternative reduces the 30-year U 3 O 8 requirement by about 9%. Additional resource utilization improvements of 10% could be realized by improved fuel management techniques. In addition to improvements achieved in reactor technology, uranium utilization can also be improved by reducing the U-235 content in the depleted uranium (tails) produced by the isotope separation facility. If the Advanced Isotope Separation Technology program, currently under development by the United States, results in a lowering of the tails assay from 0.20 w/o to 0.05 w/o the uranium feed requirement for MEU/Th cycles would be further reduced by 22%. A total improvement of 41% over the already relatively low 4280 ST/GWe net lifetime U 3 O 8 requirement would result in a 2525 ST/GWe 30-year yet U 3 O 8 requirement if all of the potential improvements were realized

  13. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes [1000 and 3000 MW(t)] and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950 0 C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950 0 C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG

  14. Advanced Natural Gas Reciprocating Engines(s)

    Energy Technology Data Exchange (ETDEWEB)

    Zurlo, James [Dresser, Inc., Addison, TX (United States)

    2012-04-05

    The ARES program was initiated in 2001 to improve the overall brake thermal efficiency of stationary, natural gas, reciprocating engines. The ARES program is a joint award that is shared by Dresser, Inc., Caterpillar and Cummins. The ARES program was divided into three phases; ARES I (achieve 44% BTE), ARES II (achieve 47% BTE) and ARES III (achieve 50% BTE). Dresser, Inc. completed ARES I in March 2005 which resulted in the commercialization of the APG1000 product line. ARES II activities were completed in September 2010 and the technology developed is currently being integrated into products. ARES III activities began in October 2010. The ARES program goal is to improve the efficiency of natural gas reciprocating engines. The ARES project is structured in three phases with higher efficiency goals in each phase. The ARES objectives are as follows: 1. Achieve 44% (ARES I), 47% (ARES II), and 50% brake thermal efficiency (BTE) as a final ARES III objective 2. Achieve 0.1 g/bhp-hr NOx emissions (with after-treatment) 3. Reduce the cost of the produced electricity by 10% 4. Improve or maintain reliability, durability and maintenance costs

  15. Empirical evaluation of the radiative cooling coefficient for krypton gas in the FTU plasma

    International Nuclear Information System (INIS)

    Fournier, K.B.; Pacella, D.; Mazzitelli, G.; Stutman, D.; Soukanovskii, V.; Goldstein, W.H.

    1997-01-01

    For future fusion reactors, a careful balance must be achieved between the cooling of the outer plasma via impurity radiation and the deleterious effects of inevitable core penetration by impurity ions. We have injected krypton gas into the Frascati Tokamak Upgrade (FTU) plasma. The measured visible bremsstrahlung and bolometric signals from krypton have been inverted and the resulting radial impurity density profile and power loss profile for krypton gas are extracted. Using the measured electron density and temperature profiles, the radiative cooling coefficient for krypton is derived. The level of intrinsic impurities (Mo, Cr, Mn and Fe) in the plasma during the krypton puffing is monitored with a VUV SPRED spectrometer. Models for krypton emissivity from the literature are compared to our measured results. 7 figs

  16. Numerical investigation of heat transfer in high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, g.; Anghaie, S. [Univ. of Florida, Gainesville, FL (United States)

    1995-09-01

    This paper proposes a computational model for analysis of flow and heat transfer in high-temperature gas-cooled reactors. The formulation of the problem is based on using the axisymmetric, thin layer Navier-Stokes equations. A hybrid implicit-explicit method based on finite volume approach is used to numerically solve the governing equations. A fast converging scheme is developed to accelerate the Gauss-Siedel iterative method for problems involving the wall heat flux boundary condition. Several cases are simulated and results of temperature and pressure distribution in the core are presented. Results of a parametric analysis for the assessment of the impact of power density on the convective heat transfer rate and wall temperature are discussed. A comparative analysis is conducted to identify the Nusselt number correlation that best fits the physical conditions of the high-temperature gas-cooled reactors.

  17. Comparison between two gas-cooled TRU burner subcritical reactors: fusion-fission and ADS

    International Nuclear Information System (INIS)

    Carluccio, T.; Rossi, P.C.R.; Angelo, G.; Maiorino, J.R.

    2011-01-01

    This work shows a preliminary comparative study between two gas cooled subcritical fast reactor as dedicated transuranics (TRU) transmuters: using a spallation neutron source or a D-T fusion neutron source based on ITER. The two concepts are compared in terms of a minor actinides burning performance. Further investigations are required to choose the best partition and transmutation strategy. Mainly due to geometric factors, the ADS shows better neutron multiplication. Other designs, like SABR and lead cooled ADS may show better performances than a Gas Coolead Subcritical Fast Reactors and should be investigated. We noticed that both designs can be utilized to transmutation. Besides the diverse source neutron spectra, we may notice that the geometric design and cycle parameters play a more important role. (author)

  18. State of development of gas cooled reactors in the Union of Soviet Socialist Republics

    International Nuclear Information System (INIS)

    Grebennik, V.N.; Mosevitskij, I.S.

    1991-01-01

    In the context of the programme for the development of gas-cooled reactors in the USSR it is reported that pilot plants with VGR-50 MW(el) and VG-400 MW(el) have been developed up to the stage of engineering design and that now the efforts are concentrated on the project of pilot-commercial reactor plant VGM (PCRP VGM) of a modular type with unit thermal power of 200-250 MW. The installation is designed to solve the main scientific and engineering problems of construction of high-temperature gas-cooled reactors, to test equipment components, and to show advantages of the given type of installations having the enhanced safety and capability to generate high-potential heat. The status of work on the PCRP VGM project is described. 3 refs, 1 fig., 1 tab

  19. Overview of environmental control aspects for the gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Nolan, A.M.

    1981-05-01

    Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included

  20. Determination of heat transfer coefficient for an interaction of sub-cooled gas and metal

    International Nuclear Information System (INIS)

    Sidek, Mohd Zaidi; Kamarudin, Muhammad Syahidan

    2016-01-01

    Heat transfer coefficient (HTC) for a hot metal surface and their surrounding is one of the need be defined parameter in hot forming process. This study has been conducted to determine the HTC for an interaction between sub-cooled gas sprayed on a hot metal surface. Both experiments and finite element have been adopted in this work. Initially, the designated experiment was conducted to obtain temperature history of spray cooling process. Then, an inverse method was adopted to calculate the HTC value before we validate in a finite element simulation model. The result shows that the heat transfer coefficient for interaction of subcooled gas and hot metal surface is 1000 W/m 2 K. (paper)

  1. Test results from a helium gas-cooled porous metal heat exchanger

    International Nuclear Information System (INIS)

    North, M.T.; Rosenfeld, J.H.; Youchison, D.L.

    1996-01-01

    A helium-cooled porous metal heat exchanger was built and tested, which successfully absorbed heat fluxes exceeding all previously tested gas-cooled designs. Helium-cooled plasma-facing components are being evaluated for fusion applications. Helium is a favorable coolant for fusion devices because it is not a plasma contaminant, it is not easily activated, and it is easily removed from the device in the event of a leak. The main drawback of gas coolants is their relatively poor thermal transport properties. This limitation can be removed through use of a highly efficient heat exchanger design. A low flow resistance porous metal heat exchanger design was developed, based on the requirements for the Faraday shield for the International Thermonuclear Experimental Reactor (ITER) device. High heat flux tests were conducted on two representative test articles at the Plasma Materials Test Facility (PMTF) at Sandia National Laboratories. Absorbed heat fluxes as high as 40 MW/m 2 were successfully removed during these tests without failure of the devices. Commercial applications for electronics cooling and other high heat flux applications are being identified

  2. Thermodynamic performance analysis of gas-fired air-cooled adiabatic absorption refrigeration systems

    International Nuclear Information System (INIS)

    Wang, L.; Chen, G.M.; Wang, Q.; Zhong, M.

    2007-01-01

    In China, the application of small size gas-fired air-cooled absorption refrigeration systems as an alternative for electric compression air conditioning systems has shown broad prospects due to occurrence of electricity peak demand in Chinese big cities and lack of water resources. However, for conventional air-cooled absorption refrigeration systems, it is difficult to enhance the heat and mass transfer process in the falling film absorber, and may cause problems, for example, remarkable increase of pressure, temperature and concentration in the generators, risk of crystallization, acceleration of corrosion, degradation of performance, and so on. This paper presents a gas-fired air-cooled adiabatic absorption refrigeration system using lithium bromide-water solutions as its working fluid, which is designed with a cooling capacity of 16 kW under standard conditions. The system has two new features of waste heat recovery of condensed water from generator and an adiabatic absorber with an air cooler. Performance simulation and characteristic analysis are crucial for the optimal control and reliability of operation in extremely hot climates. A methodology is presented to simulate thermodynamic performance of the system. The influences of outdoor air temperature on operation performances of the system are investigated

  3. Designing a gas cooled ADS for enhanced waste transmutation. The PDS-XADS European Project contribution

    International Nuclear Information System (INIS)

    Rimpault, G.; Sunderland, R.; Mueller, A.C.

    2006-01-01

    Accelerator driven system (ADS) are complex in their conception. It is the reason why studies proceed step by step. At the moment, one can take advantage of the work performed within the PDS-XADS project (Preliminary Design Studies of an eXperimental ADS) of the 5. European programme. The PDS-XADS project has been the first one to define rather detailed plants for a demonstration of the ADS technology, making a full use of European expertise from different research organizations, industries and universities. This first step was using MOX fuel technology with a design mostly devoted to the technology demonstration. Elaborated designs are sufficiently advanced to confirm the good prospects in the feasibility of such ADS plants. Also weak points have been identified and it is not a surprise that the open issues appear in the most unusual parts of reactor design i.e. in the spallation module. For what concerns the accelerator, the high reliability/availability requirements remain an important issue. The strategy to overcome these difficulties is a standard practice in reliability engineering, a technical discipline for risk estimation and management that is followed for many industrial applications or products in various fields. The gas technology exhibits clear interests in terms of coolant chemical inertness, overall simplicity of the reactor (internals, components) that can be based on proven helium cooled reactor experience but the chosen volume power (56 W/cm 3 ) for this concept is an upper limit due to constraints to the mechanical behaviour of the steel of the cladding. On the other hand, the removal of the decay heat is very much associated to the use of active systems even in protected transients i.e. with proton beam interruption. The statistical safety analysis has demonstrated however that the heat exchangers are the less reliable part of the DHR system. A solution to overcome this difficulty is the use of redundant and diversified systems. The final

  4. Optimizing parameters of GTU cycle and design values of air-gas channel in a gas turbine with cooled nozzle and rotor blades

    Science.gov (United States)

    Kler, A. M.; Zakharov, Yu. B.

    2012-09-01

    The authors have formulated the problem of joint optimization of pressure and temperature of combustion products before gas turbine, profiles of nozzle and rotor blades of gas turbine, and cooling air flow rates through nozzle and rotor blades. The article offers an original approach to optimization of profiles of gas turbine blades where the optimized profiles are presented as linear combinations of preliminarily formed basic profiles. The given examples relate to optimization of the gas turbine unit on the criterion of power efficiency at preliminary heat removal from air flows supplied for the air-gas channel cooling and without such removal.

  5. The modular high-temperature gas-cooled reactor: A cost/risk competitive nuclear option

    International Nuclear Information System (INIS)

    Gotschall, H.L.

    1994-01-01

    The business risks of nuclear plant ownership are identified as a constraint on the expanded use of nuclear power. Such risks stem from the exacting demands placed on owner/operator organizations of current plants to demonstrate ongoing compliance with safety regulations and the resulting high costs for operation and maintenance. This paper describes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) design, competitive economics, and approach to reducing the business risks of nuclear plant ownership

  6. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    Energy Technology Data Exchange (ETDEWEB)

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

  7. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    International Nuclear Information System (INIS)

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc

  8. Probabilistic safety assessment framework of pebble-bed modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Liu Tao; Tong Jiejuan; Zhao Jun; Cao Jianzhu; Zhang Liguo

    2009-01-01

    After an investigation of similar reactor type probabilistic safety assessment (PSA) framework, Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) PSA framework was presented in correlate with its own design characteristics. That is an integral framework which spreads through event sequence structure with initiating events at the beginning and source term categories in the end. The analysis shows that it is HTR-PM design feature that determines its PSA framework. (authors)

  9. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  10. A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

    International Nuclear Information System (INIS)

    Dubourg, M.

    1993-01-01

    This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable

  11. Characterization of effluents from a high-temperature gas-cooled reactor fuel refabrication plant

    International Nuclear Information System (INIS)

    Judd, M.S.; Bradley, R.A.; Olsen, A.R.

    1975-12-01

    The types and quantities of chemical and radioactive effluents that would be released from a reference fuel refabrication facility for the High-Temperature Gas-Cooled Reactor (HTGR) have been determined. This information will be used to predict the impact of such a facility on the environment, to identify areas where additional development work needs to be done to further identify and quantify effluent streams, and to limit effluent release to the environment

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  14. Frequency and distribution of leakages in steam generators of gas-cooled reactors

    International Nuclear Information System (INIS)

    Bongratz, R.; Breitbach, G.; Wolters, J.

    1988-01-01

    In gas cooled reactors with graphitic primary circuit structures - such as HTR, AGR or Magnox - the water ingress is an event of great safety concern. Water or steam entering the primary circuit react with the hot graphite and carbon-oxide and hydrogen are produced. As the most important initiating event a leak in a steam generator must be taken into account. From the safety point of view as well as for availability reasons it is necessary to construct reliable boilers. Thus the occurrence of a boiler leak should be a rare event. In the context of a probabilistic safety study for an HTR-Project much effort was invested to get information about the frequency and the size distribution of tube failures in steam generators of gas cooled reactors. The main data base was the boiler tube failure statistics of United Kingdom gas cooled reactors. The data were selected and applied to a modern HTR steam generator design. A review of the data showed that the failure frequency is not connected with the load level (pressures, temperatures) or with the geometric size of the heating surface of the boiler. Design, construction, fabrication, examination and operation conditions have the greatest influence an the failure frequency but they are practically not to be quantified. The typical leak develops from smallest size. By erosion effects of the entering water or steam it is enlarged to perhaps some mm 2 , then usually it is detected by moisture monitors. Sudden tube breaks were not reported in the investigated period. As a rule boiler leaks in gas cooled reactors are much more, rare then leaks in steam generators of light water reactors and fossil fired boilers. (author)

  15. Specialists' meeting on heat exchanging components of gas-cooled reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components

  16. Study on fundamental features of helium turbomachine for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Jie; Gu Yihua

    2004-01-01

    The High temperature gas-cooled reactor (HTGR) coupled with helium turbine cycle is considered as one of the leading candidates for future nuclear power plants. The HTGR helium turbine cycle was analyzed and optimized. Then the focal point of investigation was concentrated on the fundamental thermodynamic and aerodynamic features of helium turbomachine. As a result, a helium turbomachine is different from a general combustion gas turbine in two main design features, that is a helium turbomachine has more blade stages and shorter blade length, which are caused by the helium property and the high pressure of a closed cycle, respectively. (authors)

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

    International Nuclear Information System (INIS)

    Barsell, A.W.

    1980-05-01

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

  18. Specialists' meeting on heat exchanging components of gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1984-07-01

    The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components.

  19. Alternative Liquid Fuel Effects on Cooled Silicon Nitride Marine Gas Turbine Airfoils

    Energy Technology Data Exchange (ETDEWEB)

    Holowczak, J.

    2002-03-01

    With prior support from the Office of Naval Research, DARPA, and U.S. Department of Energy, United Technologies is developing and engine environment testing what we believe to be the first internally cooled silicon nitride ceramic turbine vane in the United States. The vanes are being developed for the FT8, an aeroderivative stationary/marine gas turbine. The current effort resulted in further manufacturing and development and prototyping by two U.S. based gas turbine grade silicon nitride component manufacturers, preliminary development of both alumina, and YTRIA based environmental barrier coatings (EBC's) and testing or ceramic vanes with an EBC coating.

  20. A design method to isothermalize the core of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takano, M.; Sawa, K.

    1987-01-01

    A practical design method is developed to isothermalize the core of block-type high-temperature gas-cooled reactors (HTGRs). Isothermalization plays an important role in increasing the design margin on fuel temperature. In this method, the fuel enrichment and the size and boron content of the burnable poison rod are determined over the core blockwise so that the axially exponential and radially flat power distribution are kept from the beginning to the end of core life. The method enables conventional HTGRs to raise the outlet gas temperature without increasing the maximum fuel temperature

  1. The main cooling system of the RSG-GAS: a study of dissabilities

    International Nuclear Information System (INIS)

    Sudiyono

    1999-01-01

    The RSG-GAS has been operating since August 1987 until now. During that period it has been many papers explain the performance and the safety features of the system. Those papers does not mean any think to the system, but the wickednesses of the system. Is still remain stick on it. This paper describes the wickednesses of the RSG-GAS main cooling system with an objective that their will be a corrective action of the management to improve the system so that such wickednesses can be minimized or even be avoided

  2. Progress Toward Buffer Gas Cooling of a 1 μB Species

    Science.gov (United States)

    Johnson, Cort; Newman, Bonna; Brahms, Nathan; Decarvalho, Robert; Li, Chih-Hao; Greytak, Tom; Kleppner, Dan; Doyle, John

    2006-05-01

    Thermalization with a non-magnetic ^3He buffer gas has been demonstrated to be an effective means of removing heat from a sample of hot magnetic atoms held in a magnetic trap[1]. To thermally isolate the atoms, it is necessary to remove the buffer gas without sweeping away the magnetically trapped atoms. The magnetic trap strength is proportional to the magnitude of the atom's moment. This is the primary reason that trapping a 1 μB species has never been realized in a buffer gas cooling experiment. Eventually, we plan to use this technique to trap and cool atomic hydrogen and deuterium. We have built a cryogenic valved cell suspended from a dilution refrigerator along the axis of a 4T quadrupole magnetic field. The cell is maintained at a base temperature of 100mK. The buffer gas is removed by opening the valve. Any remaining buffer gas may be energetic enough to scatter the atoms out of the trapping potential, thus limiting the lifetime. Therefore, we lower the temperature of the cell to decrease the vapor pressure of any ^3He remaining on the walls. We can mimic the conditions of a 1 μB species using atomic Mn (5 μB) simply by lowering the trapping field strength by a factor of five. Further, we will report on efforts to trap and thermally isolate ^7Li, a true 1 μB species. [1]J. D. Weinstein et al., Phys. Rev. A 57, R3173 (1998).

  3. Parametric study of sodium aerosols in the cover-gas space of sodium-cooled reactors

    International Nuclear Information System (INIS)

    Sheth, A.

    1975-03-01

    A mathematical model has been developed to describe the behavior of sodium aerosols in the cover-gas space of a sodium-cooled reactor. A review of the literature was first made to examine methods of aerosol generation, mathematical expressions representing aerosol behavior, and pertinent experimental investigations of sodium aerosols. In the development of the model, some terms were derived from basic principles and other terms were estimated from available correlations. The model was simulated on a computer, and important parameters were studied to determine their effects on the overall behavior of sodium aerosols. The parameters studied were sodium pool temperature, source and initial size of particles, film thickness at the sodium pool/cover gas interface, wall plating parameters, cover-gas flow rate, and type of cover gas (argon and helium). The model satisfactorily describes the behavior of sodium aerosol in argon, but not in helium. Possible reasons are given for the failure of the model with helium, and further experimental work is recommended. The mathematical model, with appropriate modifications to describe the behavior of sodium aerosols in helium, would be very useful in designing traps to remove aerosols from the cover gas of sodium-cooled reactors. (U.S.)

  4. Development of Modified Incompressible Ideal Gas Model for Natural Draft Cooling Tower Flow Simulation

    Science.gov (United States)

    Hyhlík, Tomáš

    2018-06-01

    The article deals with the development of incompressible ideal gas like model, which can be used as a part of mathematical model describing natural draft wet-cooling tower flow, heat and mass transfer. It is shown, based on the results of a complex mathematical model of natural draft wet-cooling tower flow, that behaviour of pressure, temperature and density is very similar to the case of hydrostatics of moist air, where heat and mass transfer in the fill zone must be taken into account. The behaviour inside the cooling tower is documented using density, pressure and temperature distributions. The proposed equation for the density is based on the same idea like the incompressible ideal gas model, which is only dependent on temperature, specific humidity and in this case on elevation. It is shown that normalized density difference of the density based on proposed model and density based on the nonsimplified model is in the order of 10-4. The classical incompressible ideal gas model, Boussinesq model and generalised Boussinesq model are also tested. These models show deviation in percentages.

  5. Development of Modified Incompressible Ideal Gas Model for Natural Draft Cooling Tower Flow Simulation

    Directory of Open Access Journals (Sweden)

    Hyhlík Tomáš

    2018-01-01

    Full Text Available The article deals with the development of incompressible ideal gas like model, which can be used as a part of mathematical model describing natural draft wet-cooling tower flow, heat and mass transfer. It is shown, based on the results of a complex mathematical model of natural draft wet-cooling tower flow, that behaviour of pressure, temperature and density is very similar to the case of hydrostatics of moist air, where heat and mass transfer in the fill zone must be taken into account. The behaviour inside the cooling tower is documented using density, pressure and temperature distributions. The proposed equation for the density is based on the same idea like the incompressible ideal gas model, which is only dependent on temperature, specific humidity and in this case on elevation. It is shown that normalized density difference of the density based on proposed model and density based on the nonsimplified model is in the order of 10-4. The classical incompressible ideal gas model, Boussinesq model and generalised Boussinesq model are also tested. These models show deviation in percentages.

  6. Economic analysis of multiple-module high temperature gas-cooled reactor (MHTR) nuclear power plants

    International Nuclear Information System (INIS)

    Liu Yu; Dong Yujie

    2011-01-01

    In recent years, as the increasing demand of energy all over the world, and the pressure on greenhouse emissions, there's a new opportunity for the development of nuclear energy. Modular High Temperature Gas-cooled Reactor (MHTR) received recognition for its inherent safety feature and high outlet temperature. Whether the Modular High Temperature Gas-cooled Reactor would be accepted extensively, its economy is a key point. In this paper, the methods of qualitative analysis and the method of quantitative analysis, the economic models designed by Economic Modeling Working Group (EMWG) of the Generation IV International Forum (GIF), as well as the HTR-PM's main technical features, are used to analyze the economy of the MHTR. A prediction is made on the basis of summarizing High Temperature Gas-cooled Reactor module characteristics, construction cost, total capital cost, fuel cost and operation and maintenance (O and M) cost and so on. In the following part, comparative analysis is taken measures to the economy and cost ratio of different designs, to explore the impacts of modularization and standardization on the construction of multiple-module reactor nuclear power plant. Meanwhile, the analysis is also adopted in the research of key factors such as the learning effect and yield to find out their impacts on the large scale development of MHTR. Furthermore, some reference would be provided to its wide application based on these analysis. (author)

  7. Nuclear power for coexistence with nature, high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1996-01-01

    Until this century, it is sufficient to aim at the winner of competition in human society to obtain resources, and to entrust waste to natural cleaning action. However, the expansion of social activities has been too fast, and the scale has become too large, consequently, in the next century, the expansion of social activities will be caught by the structure of trilemma that is subjected to the strong restraint and selection from the problems of finite energy and resources and environment preservation. In 21st century, the problems change to those between mankind and nature. Energy supply and population increase, envrionment preservation and human activities, and the matters that human wisdom should bear regarding energy technology are discussed. In Japan, the construction of the high temperature engineering test reactor (HTTR) is in progress. The design of high temperature gas-cooled reactors and their features on the safety are explained. The capability of reducing CO 2 release of high temperature gas-cooled reactors is reported. In future, it is expected that the time of introducing high temperature gas-cooled reactors will come. (K.I.)

  8. Advancement of Miniature Optic Gas Sensor (MOGS) Probe Technology

    Science.gov (United States)

    Chullen, Cinda

    2015-01-01

    Advancement of Miniature Optic Gas Sensor (MOGS) Probe Technology" project will investigate newly developed optic gas sensors delivered from a Small Business Innovative Research (SBIR) Phase II effort. A ventilation test rig will be designed and fabricated to test the sensors while integrated with a Suited Manikin Test Apparatus (SMTA). Once the sensors are integrated, a series of test points will be completed to verify that the sensors can withstand Advanced Suit Portable Life Support System (PLSS) environments and associated human metabolic profiles for changes in pressure and levels of Oxygen (ppO2), carbon dioxide (ppCO2), and humidity (ppH2O).

  9. A novel nuclear combined power and cooling system integrating high temperature gas-cooled reactor with ammonia–water cycle

    International Nuclear Information System (INIS)

    Luo, Chending; Zhao, Fuqiang; Zhang, Na

    2014-01-01

    Highlights: • We propose a novel nuclear ammonia–water power and cooling cogeneration system. • The high temperature reactor is inherently safe, with exhaust heat fully recovered. • The thermal performances are improved compared with nuclear combined cycle. • The base case attains an energy efficiency of 69.9% and exergy efficiency of 72.5%. • Energy conservation and emission reduction are achieved in this cogeneration way. - Abstract: A nuclear ammonia–water power and refrigeration cogeneration system (NAPR) has been proposed and analyzed in this paper. It consists of a closed high temperature gas-cooled reactor (HTGR) topping Brayton cycle and a modified ammonia water power/refrigeration combined bottoming cycle (APR). The HTGR is an inherently safe reactor, and thus could be stable, flexible and suitable for various energy supply situation, and its exhaust heat is fully recovered by the mixture of ammonia and water in the bottoming cycle. To reduce exergy losses and enhance outputs, the ammonia concentrations of the bottoming cycle working fluid are optimized in both power and refrigeration processes. With the HTGR of 200 MW thermal capacity and 900 °C/70 bar reactor-core-outlet helium, the system achieves 88.8 MW net electrical output and 9.27 MW refrigeration capacity, and also attains an energy efficiency of 69.9% and exergy efficiency of 72.5%, which are higher by 5.3%-points and 2.6%-points as compared with the nuclear combined cycle (NCC, like a conventional gas/steam power-only combined cycle while the topping cycle is a closed HTGR Brayton cycle) with the same nuclear energy input. Compared with conventional separate power and refrigeration generation systems, the fossil fuel saving (based on CH 4 ) and CO 2 emission reduction of base-case NAPR could reach ∼9.66 × 10 4 t/y and ∼26.6 × 10 4 t/y, respectively. The system integration accomplishes the safe and high-efficiency utilization of nuclear energy by power and refrigeration

  10. Economic aspects of advanced coal-fired gas turbine locomotives

    Science.gov (United States)

    Liddle, S. G.; Bonzo, B. B.; Houser, B. C.

    1983-01-01

    Increases in the price of such conventional fuels as Diesel No. 2, as well as advancements in turbine technology, have prompted the present economic assessment of coal-fired gas turbine locomotive engines. A regenerative open cycle internal combustion gas turbine engine may be used, given the development of ceramic hot section components. Otherwise, an external combustion gas turbine engine appears attractive, since although its thermal efficiency is lower than that of a Diesel engine, its fuel is far less expensive. Attention is given to such a powerplant which will use a fluidized bed coal combustor. A life cycle cost analysis yields figures that are approximately half those typical of present locomotive engines.

  11. THE DIFFERENT EVOLUTION OF GAS AND DUST IN DISKS AROUND SUN-LIKE AND COOL STARS

    International Nuclear Information System (INIS)

    Pascucci, I.; Apai, D.; Luhman, K.; Henning, Th.; Bouwman, J.; Meyer, M. R.; Lahuis, F.; Natta, A.

    2009-01-01

    Planet formation is profoundly impacted by the properties of protoplanetary disks and their central star. However, how disk properties vary with stellar parameters remains poorly known. Here, we present the first comprehensive, comparative Spitzer/IRS study of the dust and gas properties of disks around young Sun-like stars (K1-M5) and cool stars/brown dwarfs (M5-M9). The comparison of these two large samples of over 60 sources reveal major differences in the evolution of both the dust and gas components. We report the first detection of organic molecules in disks around brown dwarfs. The detection rate statistics and the line flux ratios of HCN and C 2 H 2 show a striking difference between the two samples, demonstrating a significant underabundance of HCN relative to C 2 H 2 in the disk surface of cool stars. We propose this to originate from the large difference in the UV irradiation around the two types of sources. The statistical comparison of the 10 μm silicate emission features also reveals a difference between the two samples. Cool stars and brown dwarfs show weaker features arising from more processed silicate grains in the disk atmosphere. These findings complement previous indications of flatter disk structures and longer disk lifetimes around cool stars. Our results highlight important differences in the chemical and physical evolution of protoplanetary disks as a function of stellar mass, temperature, and radiation field which should be taken into account in planet formation models. We note that the different chemistry of preplanetary materials in the disk may also influence the bulk composition and volatile content of the forming planets. In particular, if exogenous HCN has played a key role in the synthesis of prebiotic molecules on Earth as proposed, then prebiotic chemistry may unfold differently on planets around cool stars.

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

    International Nuclear Information System (INIS)

    Chang Oh

    2006-01-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 900 C and operational fuel temperatures above 1250 C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR's higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

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

  14. Impact of roughness on the instability of a free-cooling granular gas

    Science.gov (United States)

    Garzó, Vicente; Santos, Andrés; Kremer, Gilberto M.

    2018-05-01

    A linear stability analysis of the hydrodynamic equations with respect to the homogeneous cooling state is carried out to identify the conditions for stability of a granular gas of rough hard spheres. The description is based on the results for the transport coefficients derived from the Boltzmann equation for inelastic rough hard spheres [Phys. Rev. E 90, 022205 (2014), 10.1103/PhysRevE.90.022205], which take into account the complete nonlinear dependence of the transport coefficients and the cooling rate on the coefficients of normal and tangential restitution. As expected, linear stability analysis shows that a doubly degenerate transversal (shear) mode and a longitudinal ("heat") mode are unstable with respect to long enough wavelength excitations. The instability is driven by the shear mode above a certain inelasticity threshold; at larger inelasticity, however, the instability is driven by the heat mode for an inelasticity-dependent range of medium roughness. Comparison with the case of a granular gas of inelastic smooth spheres confirms previous simulation results about the dual role played by surface friction: while small and large levels of roughness make the system less unstable than the frictionless system, the opposite happens at medium roughness. On the other hand, such an intermediate window of roughness values shrinks as inelasticity increases and eventually disappears at a certain value, beyond which the rough-sphere gas is always less unstable than the smooth-sphere gas. A comparison with some preliminary simulation results shows a very good agreement for conditions of practical interest.

  15. Thermodynamic data for selected gas impurities in the primary coolant of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Feber, R.C.

    1976-12-01

    The literature of thermodynamic data for selected fission-product species is reviewed and supplemented in support of complex chemical equilibrium calculations applied to fission-product distributions in the primary coolant of high-temperature gas-cooled reactors. Thermodynamic functions and heats and free energies of formation are calculated and tabulated to 3000 0 K for CsI (s,l,g), Cs 2 I 2 (g), CH 3 I(g), COI 2 (g), and CsH(g). 79 references

  16. Nodalization Preparation for the Transient Simulation of Cooling System for One Line Mode of RSG-GAS

    International Nuclear Information System (INIS)

    Sukmanto Dibyo; Susyadi; Tagor MS; Darwis Isnaeni

    2004-01-01

    Cooling system is important component in RSG-GAS. To carry out the transient simulation of one line-cooling mode, the model of RSG-GAS has been prepared. To illustrate the transient condition, the RELAP5.MOD3 computer code the existing input files were used. This Input consist of kinetic, thermal, hydraulic and geometries data. Modification and decrement of number of nodalization has been done to simplification as well as running time. The reasonable result of model is arranged to determine the initial condition of input data therefore steady state condition have agreement to the analysis result of one line cooling mode of RSG-GAS. Parameter investigated are transient temperatures of cooling system after decreasing of secondary cooling system occur as function of time. These parameters can be requested using input of Minor Edit Request Simulation is conducted at the reactor power of 15 MW steady-state for one-line cooling mode in which the primary and secondary cooling of 430 kg/sec and 550 kg/sec respectively. Decreasing of secondary cooling flow is caused by pump trip. As a consequence, the control rod drop due to reactor protection system. The negative reactivity of control rod causes decreasing of reactor power. Change of pattern for the primary and secondary cooling system can be known. After that simulation depicts that increasing of temperatures occur at the certain moment since initiation temperature conditions, due to reactor shut down, curve inclined move going down. (author)

  17. Theoretical study and design of a low-grade heat-driven pilot ejector refrigeration machine operating with butane and isobutane and intended for cooling of gas transported in a gas-main pipeline

    KAUST Repository

    Petrenko, V.O.; Volovyk, O.S.

    2011-01-01

    This paper describes the construction and performance of a novel combined system intended for natural gas transportation and power production, and for cooling of gas transported in a gas-main pipeline. The proposed system includes a gas turbine

  18. A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1975-01-01

    The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

  19. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-04-15

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  20. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    International Nuclear Information System (INIS)

    2013-04-01

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  1. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

    The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  2. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Grodzki Marcin

    2017-12-01

    Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  3. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  4. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

    1988-01-01

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

  5. Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Hang; Wang Jie; Laurien, E.

    2010-01-01

    The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used in the geometry variation was discussed with the reference of experimental data. The original thin ribs in the current design were merged into thicker ones, and a junction located at the starting end of the outlet pipe was introduced. After comparing several potential optimization methods, an improved geometry was selected with the merged ribs increasing the pre-defined mixing coefficient and the junction reducing the pressure drop. Future work was discussed based on the simulation of real reactor case. The work shows a direction for design improvements of the lower plenum geometry. (authors)

  6. Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

    1975-06-01

    The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

  7. The cryogenic cooling program in high-heat-load optics at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Rogers, C.S.

    1993-07-01

    This paper describes some of the aspects of the cryogenic optics program at the Advanced Photon Source (APS). A liquid-nitrogen-cooled, high-vacuum, double crystal monochromator is being fabricated at Argonne National Laboratory (ANL). A pumping system capable of delivering a variable flow rate of up to 10 gallons per minute of pressurized liquid nitrogen and removing 5 kilowatts of x-ray power is also being constructed. This specialized pumping system and monochromator will be used to test the viability of cryogenically cooled, high-heat-load synchrotron optics. It has been determined that heat transfer enhancement will be required for optics used with APS insertion devices. An analysis of a porous-matrix-enhanced monochromator crystal is presented. For the particular case investigated, a heat transfer enhancement factor of 5 to 6 was calculated

  8. Perspectives on understanding and verifying the safety terrain of modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Donald E., E-mail: donald@carlsonperin.net [11221 Empire Lane, Rockville, MD 20852 (United States); Ball, Sydney J., E-mail: beckysyd@comcast.net [100 Greywood Place, Oak Ridge, TN 37830 (United States)

    2016-09-15

    The passive safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving loss of forced cooling, coolant depressurization, air ingress, moisture ingress, and reactivity events. In addition to discussing associated uncertainties and potential measures to address them, this paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs.

  9. Modular high-temperature gas-cooled reactor simulation using parallel processors

    International Nuclear Information System (INIS)

    Ball, S.J.; Conklin, J.C.

    1989-01-01

    The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to very-low-probability severe accidents. The simulations routines, which had originally been developed entirely as serial code, were readily adapted to parallel processing Fortran. The resulting parallelized simulation speed was enhanced significantly. Workstation interfaces are being developed to provide for user (operator) interaction. In this paper the benefits realized by adapting previous MHTGR codes to run on a parallel processor are discussed, along with results of typical accident analyses

  10. The design, safety and project development status of the modular high temperature gas-cooled reactor in the United States

    International Nuclear Information System (INIS)

    Mears, L.D.; Dean, R.A.

    1987-01-01

    The cooperative government and industry Modular High Temperature Gas-Cooled Reactor (MHTGR) Program in the United States has advanced a 350 MW(t) plant design through the conceptual development stage. The system incorporates an annular core of prismatic fuel elements within a steel pressure vessel connected, in a side-by-side arrangement, by a concentric duct to a second steel vessel containing a steam generator and helium coolant circulator. The reference plant design consists of four reactor modules installed in separate below-grade silos, providing steam to two conventional turbine generators. The nominal net plant output is 540 MW(e). The small reactor system takes unique advantage of the high temperature capability of the refractory coated fuel and the large thermal inertia of the graphite moderator to provide a design capable of withstanding a complete loss of active core cooling without causing excessive core heatup and significant release of fission products from the fuel. Present program activities are concentrated on interactions with the Nuclear Regulatory Commission aimed at obtaining a Licensability Statement. A project initiative to build a prototype plant which would demonstrate the MHTGR-unique licensing process, plant performance, costs and schedule plus establish an industrial infrastructure to proceed with follow-on commercial MHTGR plants by the turn of the century, is being undertaken by the utility/vendor participants (author)

  11. A literature survey on gas turbines materials - recent advances

    International Nuclear Information System (INIS)

    Gras, J.M.

    1992-10-01

    The 9001F gas turbine (rating of about 200 MW) is one of the most recent versions of the 9000 series, benefitting from the developments and technological advances, notably in regard to structural materials. In the framework of the EDF gas turbine engineering and construction program, evaluating the nature of these developments can provide guidance in appraising the construction materials proposed by other manufacturers. After a brief comparison between the Gennevilliers 9001F engine and the 85 MW 9000B gas turbine at Bouchain, ordered by EDF in 1971, various research aspects for optimizing gas turbine refractory material mechanical properties and corrosion resistance (superalloys, monolithic ceramics and composite ceramics) are presented; present current and future trends for high power equipment of this type are also discussed

  12. The absorption and emission spectrum of radiative cooling galactic fountain gas

    Science.gov (United States)

    Benjamin, Robert A.; Shapiro, Paul R.

    1993-01-01

    We have calculated the time-dependent, nonequilibrium thermal and ionization history of gas cooling radiatively from 10(exp 6) K in a one-dimensional, planar, steady-state flow model of the galactic fountain, including the effects of radiative transfer. Our previous optically thin calculations explored the effects of photoionization on such a flow and demonstrated that self-ionization was sufficient to cause the flow to match the observed galactic halo column densities of C 4, Si 4, and N 5 and UV emission from C 4 and O 3 in the constant density (isochoric) limit, which corresponded to cooling regions homogeneous on scales D less than or approximately equal to 1 kpc. Our new calculations which take full account of radiative transfer confirm the importance of self-ionization in enabling such a flow to match the data but allow a much larger range for cooling region sizes, i.e. D(sub 0) greater than or approximately equal to 15 pc. For an initial flow velocity v(sub 0) approximately equal to 100 km/s, comparable to the sound speed of a 10(exp 6) K gas, the initial density is found to be n(sub h,0) is approximately 2 x 10(exp -2) cm(exp -3), in reasonable agreement with other observation estimates, and D(sub 0) is approximately equal to 40 pc. We also compare predicted H(alpha) fluxes, UV line emission, and broadband x-ray fluxes with observed values. One dimensional numerical hydrodynamical calculations including the effects of radiative cooling are also presented.

  13. Experimental study of gas engine driven air to water heat pump in cooling mode

    International Nuclear Information System (INIS)

    Elgendy, E.; Schmidt, J.

    2010-01-01

    Nowadays a sustainable development for more efficient use of energy and protection of the environment is of increasing importance. Gas engine heat pumps represent one of the most practicable solutions which offer high energy efficiency and environmentally friendly for heating and cooling applications. In this paper, the performance characteristics of gas engine driven heat pump used in water cooling were investigated experimentally without engine heat recovery. The effects of several important factors (evaporator water inlet temperature, evaporator water volume flow rate, ambient air temperature, and engine speed) on the performance of gas engine driven heat pump were studied in a wide range of operating conditions. The results showed that primary energy ratio of the system increased by 22.5% as evaporator water inlet temperature increased from 13 o C to 24 o C. On the other hand, varying of engine speed from 1300 rpm to 1750 rpm led to decrease in system primary energy ratio by 13%. Maximum primary energy ratio has been estimated with a value of two over a wide range of operating conditions.

  14. Basic study on high temperature gas cooled reactor technology for hydrogen production

    International Nuclear Information System (INIS)

    Chang, Jong Hwa; Lee, W. J.; Lee, H. M.

    2003-01-01

    The annual production of hydrogen in the world is about 500 billion m 3 . Currently hydrogen is consumed mainly in chemical industries. However hydrogen has huge potential to be consumed in transportation sector in coming decades. Assuming that 10% of fossil energy in transportation sector is substituted by hydrogen in 2020, the hydrogen in the sector will exceed current hydrogen consumption by more than 2.5 times. Currently hydrogen is mainly produced by steam reforming of natural gas. Steam reforming process is chiefest way to produce hydrogen for mass production. In the future, hydrogen has to be produced in a way to minimize CO2 emission during its production process as well as to satisfy economic competition. One of the alternatives to produce hydrogen under such criteria is using heat source of high-temperature gas-cooled reactor. The high-temperature gas-cooled reactor represents one type of the next generation of nuclear reactors for safe and reliable operation as well as for efficient and economic generation of energy

  15. Development and Performance of an Advanced Ejector Cooling System for a Sustainable Built Environment

    Directory of Open Access Journals (Sweden)

    Paulo ePereira

    2015-06-01

    Full Text Available Ejector refrigeration is a promising technology for the integration into solar driven cooling systems because of its relative simplicity and low initial cost. The major drawback of such a system is associated to its relatively low coefficient of performance (COP under variable operating conditions. In order to overcome this problem, an advanced ejector was developed that changes its geometrical features depending on the upstream and downstream conditions. This paper provides a short overview of the development process and results of a small cooling capacity (1.5 kW solar driven cooling system using a variable geometry ejector. During the design steps, a number of theoretical works have been carried out, including the selection of the working fluid, the determination of the geometrical requirements and prototype design. Based on the analysis, R600a was selected as working fluid. A prototype was constructed with two independent variable geometrical factors: the area ratio and the nozzle exit position. A test rig was also assembled in order to test the ejector performance under controlled laboratory conditions and to elaborate a control algorithm for the variable geometry. Ejector performance was assessed by calculation of cooling cycle COP, entrainment ratio and critical back pressure. The results show that for a condenser pressure of 3 bar, an 80% increase in the COP was obtained when compared to the performance of a fixed geometry ejector. Experimental COP values varied between 0.4 and 0.8, depending on operating conditions. Currently the cooling cycle is being integrated into a solar driven demonstration site for long term in situ assessment.

  16. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

    2013-10-15

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

  17. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

  18. Status of advanced technology and design for water cooled reactors: Heavy water reactors

    International Nuclear Information System (INIS)

    1989-07-01

    In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

  19. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    1988-10-01

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  20. Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinery. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper. The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle

  1. Materials and structural aspects of advanced gas-turbine helicopter engines

    Science.gov (United States)

    Freche, J. C.; Acurio, J.

    1979-01-01

    Advances in materials, coatings, turbine cooling technology, structural and design concepts, and component-life prediction of helicopter gas-turbine-engine components are presented. Stationary parts including the inlet particle separator, the front frame, rotor tip seals, vanes and combustors and rotating components - compressor blades, disks, and turbine blades - are discussed. Advanced composite materials are considered for the front frame and compressor blades, prealloyed powder superalloys will increase strength and reduce costs of disks, the oxide dispersion strengthened alloys will have 100C higher use temperature in combustors and vanes than conventional superalloys, ceramics will provide the highest use temperature of 1400C for stator vanes and 1370C for turbine blades, and directionally solidified eutectics will afford up to 50C temperature advantage at turbine blade operating conditions. Coatings for surface protection at higher surface temperatures and design trends in turbine cooling technology are discussed. New analytical methods of life prediction such as strain gage partitioning for high temperature prediction, fatigue life, computerized prediction of oxidation resistance, and advanced techniques for estimating coating life are described.

  2. Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

    International Nuclear Information System (INIS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

    2006-10-01

    Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

  3. Stability of test environments for performance evaluation of materials for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Edgemon, G.L.; Wilson, D.F.; Bell, G.E.C.

    1993-01-01

    Stability of the primary helium-based coolant test gas for use in performance ests of materials for the Modular High-Temperature Gas-Cooled Reactor (MHTGR) was determined. Results of tests of the initial gas chemistry from General Atomics (GA) at elevated temperatures, and the associated results predicted by the SOLGASMIX trademark modelling package are presented. Results indicated that for this gas composition and at flow rates obtainable in the test loop, 466 ± 24C is the highest temperature that can be maintained without significantly altering the specified gas chemistry. Four additional gas chemistries were modelled using SOLGASMIX trademark

  4. Liquid metal versus gas cooled reactor concepts for a turbo electric powered space vehicle

    International Nuclear Information System (INIS)

    Carre, F.; Proust, E.; Schwartz, J.P.

    1985-01-01

    Recent CNES/CEA prospective studies of an orbit transfer vehicule to be launched by ARIANE V, emphasize the advantage of the Brayton cycle over the thermionics and thermoelectricity, in minimizing the total mass of 100 to 300 kWsub(e) power systems under the constraint specific to ARIANE of a radiator area limited to 95 m 2 . The review of candidate reactor concepts for this application, finally recommends both liquid metal and gas cooled reactors, for their satisfactory adaptation to a reference Brayton cycle and for the available experience from the terrestrial operation of comparable systems

  5. The Preliminary Study of High Temperature Gas Cooled Reactors (HTGRs) Technology

    International Nuclear Information System (INIS)

    Nurfarhana Ayuni Joha; Izhar Abu Hussin; Ridzuan Abdul Mutalib

    2015-01-01

    High Temperature Gas Cooled Reactors (HTGRs) have attracted worldwide interest because of their high outlet temperatures, which allow them to be used for applications beyond electricity generation. HTGRs have been built and operated since as far back as the 1970s. Experimental and demonstration reactors of this type have operated in China, Great Britain, Germany, Japan, and the United States of America. This paper is written to share the valuable knowledge and information of HTGRs technology as a mean to enrich peoples understanding of the technology. This paper will present the technological features of HTGRs that allow for a modular design with inherently safe characteristics. (author)

  6. Conceptual design study of high temperature gas-cooled reactor for plutonium incineration

    International Nuclear Information System (INIS)

    Goto, Minoru

    2013-01-01

    JAEA has started a conceptual design study of a Pu burner HTGR, which is called CBHTR (Clean Burn High Temperature gas-cooled Reactor). CBHTR’s fuel is TRISO-coated fuel particle with PuO 2 -YSZ (Yttria- Stabilized Zirconia) kernel, which increase proliferation resistance, safety of geological disposal, and Pu incineration. CBHTR can decrease Puf ratio from 60% to 20% with 520 GWd/t. In the future, 15% of electricity capacity is employed by 7 of CBHTRs and 59 of U-HTRs. JAEA has a R and D plan of manufacturing technology of TRISO-coated fuel with PuO 2 -YSZ kernel

  7. The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

    International Nuclear Information System (INIS)

    Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W.; Verfondern, K.

    1988-01-01

    High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur

  8. Preliminary study of radionuclide corrosion products in primary cooling water at RSG-GAS

    International Nuclear Information System (INIS)

    Lestari, D.E.; Pudjojanto, M.S.; Subiharto; Budi, S.

    1998-01-01

    Analysis of radionuclides emitting gamma rays at the primary cooling water at RSG-GAS has been carried out. The water coolant samples was performed using a low level background gamma spectrometer unit, including of high resolution of gamma detector HP-Ge Tennelec and Multichannel Analyzer (MCA) ADCAM 100 ORTEC. The result indicated Na-24 and Mn-56 radionuclides that may be as corrosion product and should studied deeply in the future. The expected activity concentration radionuclide for Mn-56 is lower than those written in the Safety Analysis Report (SAR), while for Na-24 is in agreement

  9. Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1985-01-01

    A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

  10. Perspectives on deployment of modular high temperature gas-cooled power plants

    International Nuclear Information System (INIS)

    Northup, T.E.; Penfield, S.

    1988-01-01

    Energy needs and energy options are undergoing re-evaluation by almost every country of the world. Energy issues such as safety, public perceptions, load growth, air pollution, acid rain, construction schedules, waste management, capital financing, project cancellations, and energy mix are but a few of those problems that are plaguing planners. This paper examines some of the key elements of the energy re-evaluation and transition that are in progress and the potential for the Modular High Temperature Gas-Cooled Reactor (Modular HTGR) to have a major impact on energy planning and its favorable prospects for deployment. (orig.)

  11. Optimization of the steam generator project of a gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Sakai, Massao

    1978-01-01

    The present work is concerned with the modeling of the primary and secondary circuits of a gas cooled nuclear reactor in order to obtain the relation between the parameters of the two cycles and the steam generator performance. The procedure allows the optimization of the steam generator, through the maximization of the plant net power, and the application of the optimal control theory of dynamic systems. The heat balances for the primary and secondary circuits are carried out simultaneously with the optimized - design parameters of the steam generator, obtained using an iterative technique. (author)

  12. Preliminary Sensitivity Study on Gas-Cooled Reactor for NHDD System Using MARS-GCR

    International Nuclear Information System (INIS)

    Lee, Seung Wook; Jeong, Jae Jun; Lee, Won Jae

    2005-01-01

    A Gas-Cooled Reactor (GCR) is considered as one of the most outstanding tools for a massive hydrogen production without CO 2 emission. Till now, two types of GCR are regarded as a viable nuclear reactor for a hydrogen production: Prismatic Modular Reactor (PMR), Pebble Bed Reactor (PBR). In this paper, a preliminary sensitivity study on two types of GCR is carried out by using MARS-GCR to find out the effect on the peak fuel and reactor pressure vessel (RPV) temperature, with varying the condition of a reactor inlet, outlet temperature, and system pressure for both PMR and PBR

  13. Safety analysis of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Mitake, Susumu; Ezaki, Masahiro; Suzuki, Katsuo; Takaya, Junichi; Shimazu, Akira

    1976-02-01

    Safety features of the experimental multi-purpose high-temperature gas-cooled reactor being developed in JAERI were studied or the basis of its preliminary conceptual design of the reactor plant. Covered are control of the plant in transients, plant behaviour in accidents, and functions of engineered safeguards, and also dynamics of the uprant and frequencies of the accidents. These studies have shown, (i) the reactor plant can be operated both in plant slave to reactor and reactor slave to plant control, (ii) stable control of

  14. Method and alloys for fabricating wrought components for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Thompson, L.D.; Johnson, W.R.

    1983-01-01

    Wrought, nickel-based alloys, suitable for components of a high-temperature gas-cooled reactor exhibit strength and excellent resistance to carburization at elevated temperatures and include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength. The range of compositions of these alloys is given. (author)

  15. Resource utilization of symbiotic high-temperature gas-cooled reactor systems

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; Brogli, R.H.

    1978-01-01

    The cumulative uranium requirements of different symbiotic combinations of high-temperature gas-cooled reactor (HTGR) prebreeders have been calculated assuming an open-end nuclear economy. The results obtained indicate that the combination of prebreeders and near-breeders does not save resources over a self-generated recycle case of comparable conversion ratio, and that it may take between 40 and 50 yr before the symbiotic system containing breeders starts saving resources over an HTGR with self-generated recycle and a conversion ratio of 0.83

  16. MHTGR [Modular High-Temperature Gas-Cooled Reactor] technology development plan

    International Nuclear Information System (INIS)

    Homan, F.J.; Neylan, A.J.

    1988-01-01

    This paper presents the approach used to define the technology program needed to support design and licensing of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The MHTGR design depends heavily on data and information developed during the past 25 years to support large HTGR (LHTGR) designs. The technology program focuses on MHTGR-specific operating and accident conditions, and on validation of models and assumptions developed using LHTGR data. The technology program is briefly outlined, and a schedule is presented for completion of technology work which is consistent with completion of a Final Safety Summary Analysis Report (FSSAR) by 1992

  17. Experimental study of gas-cooled current leads for superconducting magnets

    International Nuclear Information System (INIS)

    Warren, R.P.

    1978-04-01

    Design details and experimental test results from several design variations of the gas-cooled, copper current leads used in conjunction with the superconducting dipole magnets for ESCAR (Experimental Superconducting Accelerator Ring) are reported. Thermal acoustic oscillations, which were experienced with an initial design, were eliminated in subsequent designs by a reduction of the hydraulic diameter. The occurrence of these oscillations is in general agreement with the stability analysis of Rott but the observed gas flow dependence is not in agreement with some other recently reported results for leads operated supercritical phase coolant. An empirically determined correlation was obtained by plotting lead resistance vs. enthalpy gain of the coolant gas. The resulting family of curves can be reduced to a single line on a plot of effective resistivity vs. the product of current and cross-sectional area divided by the product of the square of the mass flow of the coolant and the lead length. This correlation, which should be applicable to other designs of copper current leads in which ideal heat transfer to the coolant gas is approached, predicts that the enthalpy gain of the coolant, and therefore the peak lead temperature, is proportional to the cube of the ratio of current to coolant mass flow. The effective value of the strongly temperature-dependent kinematic viscosity of the coolant gas was found to vary linearly with the effective resistivity of the lead

  18. Advances in measuring techniques for turbine cooling test rigs - Status report

    Science.gov (United States)

    Pollack, F. G.

    1974-01-01

    Instrumentation development pertaining to turbine cooling research has resulted in the design and testing of several new systems. Pressure measurements on rotating components are being made with a rotating system incorporating ten miniature transducers and a slip-ring assembly. The system has been tested successfully up to speeds of 9000 rpm. An advanced system development combining pressure transducer and thermocouple signals is also underway. Thermocouple measurements on rotating components are transferred off the shaft by a 72-channel rotating data system. Thermocouple data channels are electronically processed on board and then removed from the shaft in the form of a digital serial train by one winding of a rotary transformer.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-11-01

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

  20. Recovery of perchloroethylene scrubbing medium generated in the refabrication of high-temperature gas-cooled reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Judd, M.S.; Van Cleve, J.E. Jr.; Rainey, W.T. Jr.

    1976-11-01

    During the refabrication of high-temperature gas-cooled reactor (HTGR) fuel, perchloroethylene (C/sub 2/Cl/sub 4/) is used as the nonmoderating scrubbing medium to remove condensable hydrocarbons, carbon soot, and uranium-bearing particulates from the off-gas streams. The process by which the contaminated perchloroethylene is recycled is discussed.