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Sample records for steam cooled reactors

  1. Steam explosions in sodium cooled breeder reactors

    International Nuclear Information System (INIS)

    Lundell, B.

    1982-01-01

    Steam explosion is considered a physical process which transport heat from molten fuel to liquid coolant so fast that the coolant starts boiling in an explosion-like manner. The arising pressure waves transform part of the thermal energy to mechanical energy. This can stress the reactor tank and threaten its hightness. The course of the explosion has not been theoretical explained. Experimental results indicate that the probability of steam explosions in a breeder reactor is small. The efficiency of the transformation of the heat of fusion into mechanical energy in substantially lower than the theoretical maximum value. The mechanical stress from the steam explosion on the reactor tank does not seem to jeopardize its tightness. (G.B.)

  2. A reflux capsule steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Lantz, E.

    Pressurized water reactor plants at numerous sites have sustained significant leakage through their steam generators. The consequent shutdowns for repairs and replacements have damaged their economics. This experience suggests that if steam generators for liquid metal fast breeder reactors (LMFBR's) continue to be built as presently designed some of them will have similar problems. Because of their larger capital investment, the consequent damage to the economics of LMFBR's could be more serious. Reflux capsules provide a way to separate sodium from water and to reduce thermal stresses in steam generators for sodium cooled reactors. Their use would also eliminate the need for a primary heat exchanger and a secondary sodium loop pump. (author)

  3. Evolution of design of steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.; Vaidyanathan

    1997-01-01

    The first sodium cooled reactor was the experimental breeder reactor (EBR-I) in usa which was commissioned in 1951 and was incidentally the first nuclear reactor to generate electrical energy. This was followed by fast breeder reactors in USSR, UK, france, USA, japan, germany and India. The use of sodium as a coolant is due to its low moderation which helps in breeding fissile fuel from fertile materials and also its high heat transfer coefficient at comparatively low velocities. The good heat transfer properties introduce thermal stresses when there are rapid changes in the sodium temperatures. Also sodium has a chemical affinity with air and water. The steam generators for sodium cooled reactors have to allow for these novel conditions and in addition, unlike other components. Choices have to be made whether it is a recirculation type as in most fossil plants or an once through unit, the power rating, shape of the tube (straight, helical, U-tube), materials (Ferritic or austenitic), with free level of sodium or not, sodium on tube side or shell side and so on. With higher pressures and steam temperatures reheating steam after partial expansion in the turbine becomes essential as in conventional turbines. For this purpose the choice of reheating fluid viz sodium or live main steam has to be made. This paper traces the evolution of steam generator designs in the different sodium cooled reactors (chronologically) and the operation experience. 16 figs., 1 tab

  4. Linear Dynamics Model for Steam Cooled Fast Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vollmer, H

    1968-04-15

    A linear analytical dynamic model is developed for steam cooled fast power reactors. All main components of such a plant are investigated on a general though relatively simple basis. The model is distributed in those parts concerning the core but lumped as to the external plant components. Coolant is considered as compressible and treated by the actual steam law. Combined use of analogue and digital computer seems most attractive.

  5. Steam generators in indirect-cycle water-cooled reactors

    International Nuclear Information System (INIS)

    Fajeau, M.

    1976-01-01

    In the indirect cycle water-cooled nuclear reactors, the steam generators are placed between the primary circuit and the turbine. They act both as an energy transmitter and as a leaktigh barrier against fission or corrosion products. Their study is thus very important from a performance and reliability point of view. Two main types are presented here: the U-tube and the once-through steam generators [fr

  6. 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

  7. 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.

  8. Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1978-02-01

    A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

  9. Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

    International Nuclear Information System (INIS)

    Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

    2012-09-01

    selection of chemistry controls is vital for NPPs with liquid metal cooled reactors. This paper highlights principles and approaches to chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors. The recommendations on how to arrange chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors are based taking into account: - the experience with operation of fossil power industry; - secondary side water chemistry of lead-bismuth eutectics cooled nuclear reactors at submarines; - steam/water cycles of NPPs with sodium cooled fast breeders BN-350 and BN-600; - secondary water chemistry at conventional NPPs with WER, RBMK and some other reactors. (authors)

  10. The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Philpott, E.F.; Pounder, F.; Willby, C.R.

    1978-01-01

    The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

  11. 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.

  12. Materials Options of Steam Generator for Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Fu Xiaogang; Long Bin; Han Liqing; Qin Bo; Zhang Jinquan; Wang Shuxing

    2013-01-01

    Overview of the material options of steam generator for sodium-cooled fast reactors, the method to calculate the service life, the thinning of wall thickness and the sodium corrosion rate, the degradation of mechanical properties (thermal aging and decarburization) and the calculation results of theoretical models

  13. Steam-generator tube failures: world experience in water-cooled nuclear power reactors in 1974

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-01-01

    Steam-generator tube failures were reported at 25 of 59 water-cooled nuclear power reactors surveyed in 1974, compared to 11 of 49 in 1973. A summary is presented of these failures, most of which, where the cause is known, were the result of corrosion. Water chemistry control, inspection and repair procedures, and failure rates are discussed

  14. 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)

  15. 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)

  16. Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-01-01

    The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

  17. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

    International Nuclear Information System (INIS)

    Pathania, R.S.; Tatone, O.S.

    1979-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

  18. Inverted Steam Generators for Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Matal, Oldřich; Šimo, Tomáš; Matal, Oldřich Jr.

    2013-01-01

    Conclusions: Two inverted steam generators of the Czech industry provenience have still been in successful operation with no water into sodium leaks at BOR 60 (RIAR Dimitrovgrad, Russian Federation). Micromodular inverted steam generator (MMISG) since 1981 and modular inverted steam generator (MISG) since 1991. In the framework of the CP ESFR project predesign studies of 100 MW (thermal) ISG modules were performed with the consideration of MMISG and MISG design, operational and safety benefits and experience. Development of material and technology for sodium heated steam generators components reflecting contemporary domestic industrial conditions in the Czech Republic was restarted in the years 2003 to 2004 and supported in the years 2008 to 2011 by the European CP ESFR project and by the Ministry of Industry and Trade of the Czech Republic

  19. Development of a nuclear steam generator system for gas-cooled reactors for application in oil sands extraction

    International Nuclear Information System (INIS)

    Smith, J.; Hart, R.; Lazic, L.

    2009-01-01

    Canada has vast energy reserves in the Oil Sands regions of Alberta and Saskatchewan. Present extraction technologies, such as strip mining, where oil deposits are close to the surface, and Steam Assisted Gravity Drainage (SAGD) technologies for deeper deposits consume significant amounts of energy to produce the bitumen and upgraded synthetic crude oil. Studies have been performed to assess the feasibility of using nuclear reactors as primary energy sources to produce, in particular the steam required for the SAGD deeper deposit extraction process. Presently available reactors fall short of meeting the requirements, in two areas: the steam produced in a 'standard' reactor is too low in pressure and temperature for the SAGD process. Requirements can be for steam as high as 12MPa pressure with superheat; and, 'standard' reactors are too large in total output. Ideally, reactors of output in the range of 400 to 500 MWth, in modules are better suited to Oil Sands applications. The above two requirements can be met using gas-cooled reactors. Generally, newer generation gas-cooled reactors have been designed for power generation, using Brayton Cycle gas turbines run directly from the heated reactor coolant (helium). Where secondary steam is required, heat recovery steam generators have been used. In this paper, a steam generating system is described which uses the high temperature helium from the reactor directly for steam generation purposes, with sufficient quantities of steam produced to allow for SAGD steam injection, power generation using a steam turbine-generator, and with potential secondary energy supply for other purposes such as hydrogen production for upgrading, and environmental remediation processes. It is assumed that the reactors will be in one central location, run by a utility type organization, providing process steam and electricity to surrounding Oil Sands projects, so steam produced is at very high pressure (12 MPa), with superheat, in order to

  20. High-temperature gas-cooled reactor steam cycle/cogeneration application study update

    International Nuclear Information System (INIS)

    1981-09-01

    Since publication of a report on the application of a High Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration (HTGR-SC/C) plant in December of 1980, progress has continued on application related activities. In particular, a reference plant and an application identification effort has been performed, a variable cogeneration cycle balance-of-plant design was developed and an updated economic analysis was prepared. A reference HTGR-SC/C plant size of 2240 MW(t) was selected, primarily on the basis of 2240 MW(t) being in the mid-range of anticipated application needs and the availability of the design data from the 2240 MW(t) Steam Cycle/Electric generation plant design. A variable cogeneration cycle plant design was developed having the capability of operating at a range of process steam loads between the reference design load (full cogeneration) and the no process steam load condition

  1. Steam generator tube performance. Experience with water-cooled nuclear power reactors during 1985

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.

    1988-12-01

    The performance of steam generator tubes at water-cooled reactors during 1985 has been reviewed. Seventy-three of 168 reactors in the survey experienced tube degradation sufficient for the tubes to be plugged. The number of tubes plugged was 6837 or 0.28% of those in service. The leading cause of tube failure was stress corrosion cracking from the primary side. Stress corrosion cracking or intergranular attack from the secondary side and pitting were also major causes of tube failure. Unlike most previous years, fretting was a substantial problem at some reactors. Overall, corrosion continued to account for more than 80% of the defects. 20 refs

  2. Secondary coolant circuit for liquid-metal cooled reactor and steam generator for such a circuit

    International Nuclear Information System (INIS)

    Brachet, A.; Figuet, J.; Guidez, J.; Lions, N.; Traiteur, R.; Zuber, T.

    1984-01-01

    An upper buffer tank and downstream buffer tank are disposed inside the steam generators. The downstream briffer tank is annular and it surrounds and communicates with a zone of the steam generator through which the liquid metal flows towards the bottom between the exchange zone and the outlet nozzle. The pressure of the inert gas blanket in the downstream buffer volume is more important than this one in the upper buffer volume. The invention applies to fast neutron nuclear reactor cooled by sodium [fr

  3. Device for the condensation of pressurized steam and its application to the cooling of a nuclear reactor after an incident

    International Nuclear Information System (INIS)

    Dagard, P.; Couturier, M.

    1989-01-01

    This document describes an invention which relates to a device for condensation of pressurized water which is at a pressure considerably above atmospheric pressure, such as the steam produced by the steam generator of a pressurized-water nuclear reactor during the cooling of the reactor after an incident. The purpose of the invention is therefore to propose a device for the condensation of steam which is under a pressure which is considerably higher than atmospheric pressure by cooling this circulating steam as a result of contact with a heat-exchange wall which is cooled by water; such a device should be easy to install in a nuclear power plant to ensure passive cooling of the reactor, it should have a very good efficiency because of efficient heat exchangers, and it should require only a limited amount of cooling water in the equipment itself

  4. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

  5. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

    International Nuclear Information System (INIS)

    Tatone, O.S.; Meindl, P.; Taylor, G.F.

    1986-06-01

    A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

  6. Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

    International Nuclear Information System (INIS)

    Panajotov, D.P.; Gorburov, V.I.

    1989-01-01

    Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

  7. AREVA Modular Steam Cycle – High Temperature Gas-Cooled Reactor Development Progress

    International Nuclear Information System (INIS)

    Lommers, L.; Shahrokhi, F.; Southworth, F.; Mayer, J. III

    2014-01-01

    The AREVA Steam Cycle – High Temperature Gas-Cooled Reactor (SCHTGR) is a modular graphite-moderated gas-cooled reactor currently being developed to support a wide variety of applications including industrial process heat, high efficiency electricity generation, and cogeneration. It produces high temperature superheated steam which makes it a good match for many markets currently dependent on fossil fuels for process heat. Moreover, the intrinsic safety characteristics of the SC-HTGR make it uniquely qualified for collocation with large industrial process heat users which is necessary for serving these markets. The NGNP Industry Alliance has selected the AREVA SC-HTGR as the basis for future development work to support commercial HTGR deployment. This paper provides a concise description of the SC-HTGR concept, followed by a summary of recent development activities. Since this concept was introduced, ongoing design activities have focused primarily on confirming key system capabilities and the suitability for potential future markets. These evaluations continue to confirm the suitability of the SC-HTGR for a variety of potential applications that are currently dependent on fossil fuels. (author)

  8. A system for regulating the pressure of resuperheated steam in high temperature gas-cooled reactor power stations

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegines, K.O.

    1975-01-01

    The invention relates to a system for regulating steam-pressure in the re-superheating portion of a steam-boiler receiving heat from a gas-cooled high temperature nuclear reactor, provided with gas distributing pumps driven by steam-turbines. The system comprises means for generating a pressure signal of desired magnitude for the re-superheating portion, and means for providing a real pressure in the re-superheating portion, means (including a by-passing device) for generating steam-flow rate signal of desired magnitude, a turbine by-pass device comprising a by-pass tapping means for regulating the steam-flow-rate in said turbine according to the desired steam-flow rate signal and means for controlling said by-pass tapping means according to said desired steam-flow-rate signal [fr

  9. Steam generator tube performance: world experience with water-cooled nuclear power reactors during 1979

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1981-01-01

    The performance of steam generator tubes in water-cooled nuclear power reactors is reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The defect rate, although higher than that in 1978, was still lower than the rates of the two previous years. Methods being employed to detect defects include the increased use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failure by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. 10 tables

  10. 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)

  11. Economic evaluation of the steam-cycle high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    1983-07-01

    The High Temperature Gas-Cooled Reactor is unique among current nuclear technologies in its ability to generate energy in temperature regimes previously limited to fossil fuels. As a result, it can offer commercial benefits in the production of electricity, and at the same time, expand the role of nuclear energy to the production of process heat. This report provides an evaluation of the HTGR-Steam Cycle (SC) system for the production of baseloaded electricity, as well as cogenerated electricity and process steam. In each case the HTGR-SC system has been evaluated against appropriate competing technologies. The computer code which was developed for this evaluation can be used to present the analyses on a cost of production or cash flow basis; thereby, presenting consistent results to a utility, interested in production costs, or an industrial steam user or third party investor, interested in returns on equity. Basically, there are two economic evaluation methodologies which can be used in the analysis of a project: (1) minimum revenue requirements, and (2) discounted cash flow

  12. Preliminary evaluation of steam generator tube rupture (SGTR) accident in lead cooled reactor

    International Nuclear Information System (INIS)

    Frano, R. Lo; Forasassi, G.

    2009-01-01

    In this paper some contributions are provided to the development of a European Lead-cooled System, known as the ELSY project (within EU-6 Framework Project); that will constitute a possible reference system for a large lead-cooled reactor of GEN IV. Steam generator (SG) tubing of this system type might be subject to a variety of degradation processes, such as cracking, wall thinning and potential leakage or rupture, eventually leading to the failure of one or more SG tubes that constitute a steam generator tube rupture (SGTR) accident with possible consequences for the safety of the primary systems. It is therefore of interest for the designer to know how the SG itself, as well as the vessel and internals structures, behave under impulsive loading conditions (in form of a rapid and strong increase of pressure) that can arise as consequences of the interaction between the primary and secondary coolants (lead-water interaction). The analysed initiator event, as already mentioned, is a large break (up to a double ended guillotine break) of one (or more) SG cooling tubes that may become severe enough to determine dangerous effects on the interested structures. In order to better simulate and perform the mentioned postulated SGTR accident sequence analyses, an appropriate numerical model with the available computing resources (FEM codes) was set up at the DIMNP of Pisa University. That model was used to evaluate the effects of the propagation of the blast pressure waves inside the SG structures, taking into account also the sloshing phenomenon that could be induced by the lead primary coolant motions. Therefore the SGTR effects study may be considered as a transient and non linear problem the solution of which provides the 'time histories' of hydrodynamic pressures and stresses on the reactor pressure vessel and internals walls. (author)

  13. Digital simulation of a commercial scale high temperature gas-cooled reactor (HTGR) steam power plant

    International Nuclear Information System (INIS)

    Ray, A.; Bowman, H.F.

    1978-01-01

    A nonlinear dynamic model of a commercial scale high temperature gas-cooled reactor (HTGR) steam power plant was derived in state-space form from fundamental principles. The plant model is 40th order, time-invariant, deterministic and continuous-time. Numerical results were obtained by digital simulation. Steady-state performance of the nonlinear model was verified with plant heat balance data at 100, 75 and 50 percent load levels. Local stability, controllability and observability were examined in this range using standard linear algorithms. Transfer function matrices for the linearized models were also obtained. Transient response characteristics of 6 system variables for independent step distrubances in 2 different input variables are presented as typical results

  14. Systems Analysis of a Fast Steam-Cooled Reactor of 1000 MW(E)

    Energy Technology Data Exchange (ETDEWEB)

    Smidt, D.; Frisch, W.; Hofmann, F.; Moers, H.; Schramm, K.; Spilker, H. [Institut fuer Reaktorentwicklung, Kernforschungszentrum, Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Kiefhaber, E. [Institut fuer Neutronenphysik und Reaktortechnik Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

    1968-05-15

    The Karlsruhe design of a steam-cooled fast reactor (Dl) has been the subject of a systems analysis. Here the dependence of fuel inventory, breeding ratio, rating, core geometry and plant efficiency on coolant pressure, and coolant temperature has been studied for two different rod powers. The effect of artificial surface roughness has been investigated. For some configurations the resulting fuel-cycle and capital costs have been determined and discussed. The main influence results from pressure. The lower pressure allows for higher breeding ratios, but lower efficiencies and vice versa. From this the fuel-cycle costs show an optimum at around 150 atm abs. The capital costs on the other side decrease with pressure. The over-all optimum of the power generating costs for the presently studied parameter range is at about 170 atm abs., a coolant outlet temperature of 540 Degree-Sign C and a rod power of 420 W/cm. Artificial roughness (boundary layer type) leads for a required system pressure and outlet temperature to a larger coolant volume fraction and, therefore, to reduced breeding ratios but higher efficiencies. As another part of the work some stability characteristics of the cores were studied. The dependence of the core stability on the varied parameters is shown. (author)

  15. Steam generator materials performance in high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Chafey, J.E.; Roberts, D.I.

    1980-11-01

    This paper reviews the materials technology aspects of steam generators for HTGRs which feature a graphite-moderated, uranium-thorium, all-ceramic core and utilizes high-pressure helium as the primary coolant. The steam generators are exposed to gas-side temperatures approaching 760 0 C and produce superheated steam at 538 0 C and 16.5 MPa (2400 psi). The prototype Peach Bottom I 40-MW(e) HTGR was operated for 1349 EFPD over 7 years. Examination after decommissioning of the U-tube steam generators and other components showed the steam generators to be in very satisfactory condition. The 330-MW(e) Fort St. Vrain HTGR, now in the final stages of startup, has achieved 70% power and generated more than 1.5 x 10 6 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, requiring a number of new materials factors including creep-fatigue and water chemistry control. Current designs of larger HTGRs also feature steam generators of helical once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers (ASME) Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of large forgings, etc

  16. Steam-generator tube failures: world experience in water-cooled nuclear power reactors during 1972

    International Nuclear Information System (INIS)

    Stevens-Guille, P.D.

    1975-01-01

    During 1972, approximately one in three operating reactors with steam generators incurred tube failures, predominantly near the tube sheet and in the bend region. Various forms of corrosion were the most frequent cause of failure. Eddy-current inspection was the preferred method for locating and investigating the cause of failure. Extensive use was made of both mechanical and explosive plugs for repair. As a class, steam generators with Monel 400 tubes had the lowest failure rates, and those with Inconel 600 tubes had the highest. (U.S.)

  17. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  18. Steam generator tube failures: world experience in water-cooled nuclear power reactors in 1975

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-11-01

    Steam generator tube failures were reported in 22 out of 62 water-cooled nuclear power plants surveyed in 1975. This was less than in 1974, and the number of the tubes affected was noticeably less. This report summarizes these failures, most of which were due to corrosion. Secondary-water chemistry control, procedures for inspection and repair, tube materials, and failure rates are discussed. (author)

  19. 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)

  20. ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

    International Nuclear Information System (INIS)

    Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G.

    1977-01-01

    ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

  1. High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

  2. 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

  3. Calculation of mass flow and steam quality distribution on fuel elements of light-water cooled boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Hermanns, H.J.

    1977-04-01

    By the example of light-water cooled nuclear reactors, the state of the calculation methods at disposal for calculating mass flow and steam quality distribution (sub-channel analysis) is indicated. Particular regard was paid to the transport phenomena occurring in reactor fuel elements in the range of two phase flow. Experimentally determined values were compared with recalculations of these experiments with the sub-channel code COBRA; from the results of these comparing calculations, conclusions could be drawn on the suitability of this code for defined applications. Limits of reliability could be determined to some extent. Based on the experience gained and the study of individual physical model concepts, recognized as being important, a sub-channel model was drawn up and the corresponding numerical computer code (SIEWAS) worked out. Experiments made at GE could be reproduced with the code SIEWAS with sufficient accuracy. (orig.) [de

  4. Thermal analysis experiment for elucidating sodium-water chemical reaction mechanism in steam generator of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki

    2012-01-01

    For the purpose of elucidating the mechanism of the sodium-water surface reaction in a steam generator of sodium-cooled fast reactors, kinetic study of the sodium (Na)-sodium hydroxide (NaOH) reaction has been carried out by using Differential Thermal Analysis (DTA) technique. The parameters, including melting points of Na and NaOH, phase transition temperature of NaOH, Na-NaOH reaction temperature, and decomposition temperature of sodium hydride (NaH) have been identified from DTA curves. Based on the measured reaction temperature, rate constant of sodium monoxide (Na 2 O) generation was obtained. Thermal analysis results indicated that Na 2 O generation at the secondary overall reaction should be considered during the sodium-water reaction. (author)

  5. Review of the cost estimate and schedule for the 2240-MWt high-temperature gas-cooled reactor steam-cycle/cogeneration lead plant

    International Nuclear Information System (INIS)

    1983-09-01

    This report documents Bechtel's review of the cost estimate and schedule for the 2240 MWt High Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration (HTGR-SC/C) Lead Plant. The overall objective of the review is to verify that the 1982 update of the cost estimate and schedule for the Lead Plant are reasonable and consistent with current power plant experience

  6. Computational methodology of sodium-water reaction phenomenon in steam generator of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Takata, Takashi; Yamaguchi, Akira; Uchibori, Akihiro; Ohshima, Hiroyuki

    2009-01-01

    A new computational methodology of sodium-water reaction (SWR), which occurs in a steam generator of a liquid-sodium-cooled fast reactor when a heat transfer tube in the steam generator fails, has been developed considering multidimensional and multiphysics thermal hydraulics. Two kinds of reaction models are proposed in accordance with a phase of sodium as a reactant. One is the surface reaction model in which water vapor reacts directly with liquid sodium at the interface between the liquid sodium and the water vapor. The reaction heat will lead to a vigorous evaporation of liquid sodium, resulting in a reaction of gas-phase sodium. This is designated as the gas-phase reaction model. These two models are coupled with a multidimensional, multicomponent gas, and multiphase thermal hydraulics simulation method with compressibility (named the 'SERAPHIM' code). Using the present methodology, a numerical investigation of the SWR under a pin-bundle configuration (a benchmark analysis of the SWAT-1R experiment) has been carried out. As a result, the maximum gas temperature of approximately 1,300degC is predicted stably, which lies within the range of previous experimental observations. It is also demonstrated that the maximum temperature of the mass weighted average in the analysis agrees reasonably well with the experimental result measured by thermocouples. The present methodology will be promising to establish a theoretical and mechanical modeling of secondary failure propagation of heat transfer tubes due to such as an overheating rupture and a wastage. (author)

  7. Cooling nuclear reactor fuel

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1975-01-01

    Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

  8. Steam up over reactor policy

    International Nuclear Information System (INIS)

    Kovan, D.

    1976-01-01

    Britain is once more assessing its nuclear power programme in the light of recent forecasts that there is unlikely to be any growth in the demand for electricity for many years to come. This means that the extra costs of launching a commercially unproven reactor, the Steam Generating Heavy Water Reactor (SGHWR), will be an even greater burden than previously expected, because they would be spread over fewer reactors. Sir John Hill's reported assessment concludes that the present strategy would be the most expensive way of developing Britain's nuclear power programme; and under the circumstances, may not be the best option. The SGHWR programme will certainly be more expensive than either relaunching a programme of advanced gas-cooled reactors (AGRs), or building American designed pressurised water reactors (PWRs). Recent developments of the AGR and PWR's and their advantages in the present position are outlined. (U.K.)

  9. Steam content of the two-phase flow in the Vk-50 boiling water cooled reactor draught section

    International Nuclear Information System (INIS)

    Fedulin, V.N.; Shmelev, V.E.; Solodkij, V.A.; Bartolomej, G.G.

    1983-01-01

    Results are presented of experimental investigation of the two-phase steam-water coolant flow hydrodynamics within the VK-50 reactor draught section. On the basis of the analysis of the obtained data a two-phase coolant flow model in a large diameter channel is proposed. It is shown that the steam-content distribution in the volume of the draught section has a pronounced non-equilibrium character manifested in the steam migration from the periphery to the central region. A minimum value of the steam content at the periphery is attained at the 0.7-1.0 m height; it is followed by a partial steam content levelling over the section. However the total steam content levelling over the cross section of the draught section does not take place. The steam distribution in the water layer over the draught section (overflow zone) is also nonuniform over the reactor section. The non-uniform steam distribution enchances with reduction nn pressure

  10. Closed loop steam cooled airfoil

    Science.gov (United States)

    Widrig, Scott M.; Rudolph, Ronald J.; Wagner, Gregg P.

    2006-04-18

    An airfoil, a method of manufacturing an airfoil, and a system for cooling an airfoil is provided. The cooling system can be used with an airfoil located in the first stages of a combustion turbine within a combined cycle power generation plant and involves flowing closed loop steam through a pin array set within an airfoil. The airfoil can comprise a cavity having a cooling chamber bounded by an interior wall and an exterior wall so that steam can enter the cavity, pass through the pin array, and then return to the cavity to thereby cool the airfoil. The method of manufacturing an airfoil can include a type of lost wax investment casting process in which a pin array is cast into an airfoil to form a cooling chamber.

  11. Accident alarm in steam generators in sodium cooled fast reactor power plants. II

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.; Taraba, O.; Hanke, V.

    1978-01-01

    Conditions were simulated in the economizer of a steam generator of water leaks in sodium at a sodium flow of O.62x10 -3 to 1.24x10 -3 m 3 /s and a sodium temperature of 320 to 380 degC by injecting water at a pressure of 6 to 10 MPa which roughly corresponds to conditions in an economizer of an actual steam generator with leaks within the limits of 0.01 to 0.3 g/s. The leak was recorded by acoustic detectors at all observed sodium flow rates and temperatures. The mean signal-to-noise ratio was in all cases greater than 2. At the assumed 25 dB noise level of the real steam generator of micromodular design it may be assumed that using existing acoustic detectors with waveguides a 0.02 g/s leak of water into sodium may be detected. The measurements showed that the technical standard of the equipment is at least as good as that of the flowmeter system of accident monitoring. (J.B.)

  12. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

    Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

  13. 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)

  14. High-temperature gas-cooled reactor steam cycle/cogeneration: lead project strategy plan

    International Nuclear Information System (INIS)

    1982-07-01

    The strategy, contained herein, for developing the HTGR system and introducing it into the energy marketplace is based on using the most developed technology path to establish a HTGR-Steam Cycle/Cogeneration (SC/C) Lead Project. Given the status of the HTGR-SC/C technology, a Lead Plant could be completed and operational by the mid 1990s. While there is remaining design and technology development that must be accomplished to fulfill technical and licensing requirements for a Lead Project commitment, the major barriers to the realization a HTGR-SC/C Lead Project are institutional in nature, e.g. budget priorities and constraints, cost/risk sharing between the public and private sector, Project organization and management, and Project financing. These problems are further complicated by the overall pervading issues of economic and regulatory instability that presently confront the utility and nuclear industries. This document addresses the major institutional issues associated with the HTGR-SC/C Lead Project and provides a starting point for discussions between prospective Lead Project participants toward the realization of such a Project

  15. The study on water ingress mass in the steam generator heat-exchange tube rupture accident of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Shi Lei; Li Fu; Zheng Yanhua

    2012-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident is an important and particular accident which will result in water ingress to the primary loop of reactor. Water ingress will result in chemical reaction of graphite fuel and structure with water, which may cause overpressure due to generation of explosive gaseous in large quantity. The study on the water ingress accident is significant for the verification of the inherent characteristics of high temperature gas-cooled reactor. The previous research shows that the amount of water ingress mass is the dominant key factor on the severity of the accident consequence. The 200 MWe high temperature gas-cooled reactor (HTR-PM), which is the first modular pebble-bed high temperature gas-cooled reactor in China designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected to be analyzed in this paper. The different DBA accident scenarios of double-ended break of single heat-exchange tube are simulated respectively by the thermal-hydraulic analysis code RETRAN-02. The results show the water ingress mass through the broken heat-exchange tube is related to the break location. The amount of water ingress mass is affected obviously by the capacity of the emptier system. With the balance of safety and economical efficiency, the amount of water ingress mass from the secondary side of steam generator into the primary coolant loop will be reduced by increasing properly the diameter of the draining lines. (authors)

  16. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Arai, Kenji.

    1996-01-01

    Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

  17. Numerical analysis of performance of steam reformer of methane reforming hydrogen production system connected with high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yin Huaqiang; Jiang Shengyao; Zhang Youjie

    2007-01-01

    Methane conversion rate and hydrogen output are important performance indexes of the steam reformer. The paper presents numerical analysis of performance of the reformer connected with high-temperature gas-cooled reactor HTR-10. Setting helium inlet flow rate fixed, performance of the reformer was examined with different helium inlet temperature, pressure, different process gas temperature, pressure, flow rate, and different steam to carbon ratio. As the range concerned, helium inlet temperature has remarkable influence on the performance, and helium inlet temperature, process gas temperature and pressure have little influence on the performance, and improving process gas flow rate, methane conversion rate decreases and hydrogen output increases, however improving steam to carbon ratio has reverse influence on the performance. (authors)

  18. Discussion on amount of water ingress mass in steam generator heat-exchange tube rupture accident of high- temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Zheng Yanhua; Shi Lei; Li Fu; Sun Ximing

    2009-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident which will result in the water ingress to the primary circuit of reactor is an important and particular accident for high-temperature gas-cooled reactor (HTGR). The analysis of the water ingress accident is significant for verifying the inherent safety characteristics of HTGR. The amount of water ingress mass is one of the decisive factors for the seriousness of the accident consequence. The 250 MW Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) designed by Institute of Nuclear and New Energy Technology of Tsinghua University was selected as an example of analysis. The analysis results show that the amount of water ingress mass is not only affected directly with the broken position and the broken area of the tubes, but also related with the diameter of draining piping and restrictor, draining control valve, action setting of emptier system. With reasonable parameters chosen, the water in steam generator could be drained effectively, so it will prevent the primary circuit of reactor from water ingress in large quantity and reduce the radioactive isotopes ingress to the secondary circuit. (authors)

  19. Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-08-15

    The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  20. Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Sang June Ahn

    2016-08-01

    Full Text Available The prototype generation IV sodium-cooled fast reactor (PGSFR has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS and the safety of the primary heat-transfer system (PHTS. In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  1. Variable electricity and steam from salt, helium and sodium cooled base-load reactors with gas turbines and heat storage - 15115

    International Nuclear Information System (INIS)

    Forsberg, C.; McDaniel, P.; Zohuri, B.

    2015-01-01

    Advances in utility natural-gas-fired air-Brayton combed cycle technology is creating the option of coupling salt-, helium-, and sodium-cooled nuclear reactors to Nuclear air-Brayton Combined Cycle (NACC) power systems. NACC may enable a zero-carbon electricity grid and improve nuclear power economics by enabling variable electricity output with base-load nuclear reactor operations. Variable electricity output enables selling more electricity at times of high prices that increases plant revenue. Peak power is achieved using stored heat or auxiliary fuel (natural gas, bio-fuels, hydrogen). A typical NACC cycle includes air compression, heating compressed air using nuclear heat and a heat exchanger, sending air through a turbine to produce electricity, reheating compressed air, sending air through a second turbine, and exhausting to a heat recovery steam generator (HRSG). In the HRSG, warm air produces steam that is used to produce added electricity. For peak power production, auxiliary heat (natural gas, stored heat) is added before the air enters the second turbine to raise air temperatures and power output. Like all combined cycle plants, water cooling requirements are dramatically reduced relative to other power cycles because much of the heat rejection is in the form of hot air. (authors)

  2. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Akiba, Miyuki.

    1996-01-01

    In a cooling device for a reactor container, a low pressure vessel is connected to an incondensible gas vent tube by way of an opening/closing valve. Upon occurrence of a loss of coolant accident, among steams and incondensible gases contained in the reactor container, steams are cooled and condensed in a heat exchanger. The incondensible gases are at first discharged from the heat exchanger to a suppression pool by way of the incondensible gas vent tube, but subsequently, they are stagnated in the incondensible gas vent tube to hinder heat exchanging and steam cooling and condensing effects in the heat exchanger thereby raising temperature and pressure in the reactor. However, if the opening/closing valve is opened when the incondensible gases are stagnated in the incondensible gas vent tube, since the incondensible gases stagnated in the heat exchanger are sucked and discharged to the low pressure vessel, the performance of the heat exchanger is maintained satisfactorily thereby enabling to suppress elevation of temperature and pressure in the reactor container. (N.H.)

  3. Emergency cooling of presurized water reactor

    International Nuclear Information System (INIS)

    Sykora, D.

    1981-01-01

    The method described of emergency core cooling in the pressurized water reactor is characterized by the fact that water is transported to the disturbed primary circuit or direct to the reactor by the action of the energy and mass of the steam and/or liquid phase of the secondary circuit coolant, which during emergency core cooling becomes an emergency cooling medium. (B.S.)

  4. 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)

  5. Reactor core cooling device

    International Nuclear Information System (INIS)

    Kobayashi, Masahiro.

    1986-01-01

    Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

  6. The Steam Generating Heavy Water Reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

  7. High-temperature gas-cooled reactor steam-cycle/cogeneration lead plant. Plant Protection and Instrumentation System design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Plant Protection and Instrumentation System provides plant safety system sense and command features, actuation of plant safety system execute features, preventive features which maintain safety system integrity, and safety-related instrumentation which monitors the plant and its safety systems. The primary function of the Plant Protection and Instrumentation system is to sense plant process variables to detect abnormal plant conditions and to provide input to actuation devices directly controlling equipment required to mitigate the consequences of design basis events to protect the public health and safety. The secondary functions of the Plant Protection and Instrumentation System are to provide plant preventive features, sybsystems that monitor plant safety systems status, subsystems that monitor the plant under normal operating and accident conditions, safety-related controls which allow control of reactor shutdown and cooling from a remote shutdown area

  8. Emergency reactor cooling device

    International Nuclear Information System (INIS)

    Arakawa, Ken.

    1993-01-01

    An emergency nuclear reactor cooling device comprises a water reservoir, emergency core cooling water pipelines having one end connected to a water feeding sparger, fire extinguishing facility pipelines, cooling water pressurizing pumps, a diesel driving machine for driving the pumps and a battery. In a water reservoir, cooling water is stored by an amount required for cooling the reactor upon emergency and for fire extinguishing, and fire extinguishing facility pipelines connecting the water reservoir and the fire extinguishing facility are in communication with the emergency core cooling water pipelines connected to the water feeding sparger by system connection pipelines. Pumps are operated by a diesel power generator to introduce cooling water from the reservoir to the emergency core cooling water pipelines. Then, even in a case where AC electric power source is entirely lost and the emergency core cooling system can not be used, the diesel driving machine is operated using an exclusive battery, thereby enabling to inject cooling water from the water reservoir to a reactor pressure vessel and a reactor container by the diesel drive pump. (N.H.)

  9. Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

    International Nuclear Information System (INIS)

    Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

    2003-01-01

    The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold

  10. 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.)

  11. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

    In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

  12. 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

  13. Nuclear reactor steam depressurization valve

    International Nuclear Information System (INIS)

    Moore, G.L.

    1991-01-01

    This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

  14. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

    The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

  15. User's instructions for ORCENT II: a digital computer program for the analysis of steam turbine cycles supplied by light-water-cooled reactors

    International Nuclear Information System (INIS)

    Fuller, L.C.

    1979-02-01

    The ORCENT-II digital computer program will perform calculations at valves-wide-open design conditions, maximum guaranteed rating conditions, and an approximation of part-load conditions for steam turbine cycles supplied with throttle steam characteristic of contemporary light-water reactors. Turbine performance calculations are based on a method published by the General Electric Company. Output includes all information normally shown on a turbine-cycle heat balance diagram. The program is written in FORTRAN IV for the IBM System 360 digital computers at the Oak Ridge National Laboratory

  16. User's instructions for ORCENT II: a digital computer program for the analysis of steam turbine cycles supplied by light-water-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, L.C.

    1979-02-01

    The ORCENT-II digital computer program will perform calculations at valves-wide-open design conditions, maximum guaranteed rating conditions, and an approximation of part-load conditions for steam turbine cycles supplied with throttle steam characteristic of contemporary light-water reactors. Turbine performance calculations are based on a method published by the General Electric Company. Output includes all information normally shown on a turbine-cycle heat balance diagram. The program is written in FORTRAN IV for the IBM System 360 digital computers at the Oak Ridge National Laboratory.

  17. Passive vibro-acoustic detection of a sodium-water reaction in a steam generator of a sodium-cooled fast neutrons nuclear reactor by beam forming

    International Nuclear Information System (INIS)

    Moriot, Jeremy

    2013-01-01

    This thesis deals with a new method to detect a sodium-water reaction in a steam generator of a fast sodium-cooled nuclear reactor. More precisely, the objective is to detect a micro-leak of water (flow ≤ 1 g/s) in less than 10 seconds by measuring the external shell vibrations of the component. The strong background noise in operation makes impossible the use of a detection system based on a threshold overrun. A beam forming method applied to vibrations measured by a linear array of accelerometers is developed in this thesis to increase the signal-to-noise ratio and to detect and locate the leak in the steam generator. A numerical study is first realized. Two models are developed in order to simulate the signals measured by the accelerometers of the array. The performances of the beam forming are then studied in function of several parameters, such as the source location and frequency, the damping factor, the background noise considered. The first model consists in an infinite plate in contact with a heavy fluid, excited by an acoustic monopole located in this fluid. Analyzing the transverse displacements in the wavenumber domain is useful to establish a criterion to sample correctly the vibration field of the plate. A second model, more representative of the system is also proposed. In this model, an elastic infinite cylindrical shell, filled with a heavy fluid is considered. The finite dimensions in the radial and circumferential directions lead to a modal behavior of the system which impacts the beam forming. Finally, the method is tested on an experimental mock-up which consists in a cylindrical pipe made in stainless steel and filled with water connected to hydraulic circuit. The water flow speed can be controlled by varying the speed of the pump. The acoustic source is generated by a hydro-phone. The performances of the beam forming are studied for different water flow speeds and different amplitude and frequencies of the source. (author) [fr

  18. 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

  19. Reactor cooling system

    International Nuclear Information System (INIS)

    Kato, Etsuji.

    1979-01-01

    Purpose: To eliminate cleaning steps in the pipelines upon reactor shut-down by connecting a filtrating and desalting device to the cooling system to thereby always clean up the water in the pipelines. Constitution: A filtrating and desalting device is connected to the pipelines in the cooling system by way of drain valves and a check valve. Desalted water is taken out from the exit of the filtrating and desalting device and injected to one end of the cooling system pipelines by way of the drain valve and the check valve and then returned by way of another drain valve to the desalting device. Water in the pipelines is thus always desalted and the cleaning step in the pipelines is no more required in the shut-down. (Kawakami, Y.)

  20. Emergency reactor cooling circuit

    International Nuclear Information System (INIS)

    Araki, Hidefumi; Matsumoto, Tomoyuki; Kataoka, Yoshiyuki.

    1994-01-01

    Cooling water in a gravitationally dropping water reservoir is injected into a reactor pressure vessel passing through a pipeline upon occurrence of emergency. The pipeline is inclined downwardly having one end thereof being in communication with the pressure vessel. During normal operation, the cooling water in the upper portion of the inclined pipeline is heated by convection heat transfer from the communication portion with the pressure vessel. On the other hand, cooling water present at a position lower than the communication portion forms cooling water lumps. Accordingly, temperature stratification layers are formed in the inclined pipeline. Therefore, temperature rise of water in a vertical pipeline connected to the inclined pipeline is small. With such a constitution, the amount of heat lost from the pressure vessel by way of the water injection pipeline is reduced. Further, there is no worry that cooling water to be injected upon occurrence of emergency is boiled under reduced pressure in the injection pipeline to delay the depressurization of the pressure vessel. (I.N.)

  1. Nuclear reactor cooling device

    International Nuclear Information System (INIS)

    Hoshi, Masaya; Makihara, Yoshiaki.

    1985-01-01

    Purpose: To improve the heat transfer performance, as well as reducing and simplifying the structure while preventing the intrusion of primary coolants to utilization systems. Constitution: Heat transfer from the primary coolant circuit to the utilization circuits is conducted by means of heat pipe type heat exchangers. The heat exchanger comprises a tightly closed vessel divided by a partition wall, through which a plurality of heat pipes are passed. The primary coolants receiving the heat from the nuclear reactor enter the first chamber of the heat exchanger to heat the evaporating portion of the heat pipes. The heated flow of steams in the heat pipes transfer to the condensating portion in the second chamber to conduct heat exchange with the utilization system. In this way, since secondary coolant circuits are saved, the heat transfer performance can be improved significantly and the risk of failure can be reduced. (Kamimura, M,)

  2. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

    2011-01-01

    Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

  3. Emergency cooling system for a liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Murata, Ryoichi; Fujiwara, Toshikatsu.

    1980-01-01

    Purpose: To suitably cool liquid metal as coolant in emergency in a liquid metal cooled reactor by providing a detector for the pressure loss of the liquid metal passing through a cooling device in a loop in which the liquid metal is flowed and communicating the detector with a coolant flow regulator. Constitution: A nuclear reactor is stopped in nuclear reaction by control element or the like in emergency. If decay heat is continuously generated for a while and secondary coolant is insufficiently cooled with water or steam flowed through a steam and water loop, a cooler is started. That is, low temperature air is supplied by a blower through an inlet damper to the cooler to cool the secondary coolant flowed into the cooler through a bypass pipe so as to finally safely stop an entire plant. Since the liquid metal is altered in its physical properties by the temperature at this time, it is detected to regulate the opening of the valve of the damper according to the detected value. (Sekiya, K.)

  4. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  5. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  6. Steam explosions in light water reactors

    International Nuclear Information System (INIS)

    1981-01-01

    The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

  7. 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

  8. 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.)

  9. 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

  10. Device for extracting steam or gas from the primary coolant line leading from a reactor pressure vessel to a straight through boiler or from the top primary boiler chamber of a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Schatz, K.

    1982-01-01

    In such a nuclear reactor, a steam or gas cushion can form when the primary system is refilled, which can cause blocking of the natural circulation or filling of the system in the area of the hot primary coolant pipe or in the top primary boiler chamber. In order to remove such a steam or gas cushion, a ventilation pipe starting from the bend of the primary coolant line is connected to the feed pipe for introducing water into the primary system. The feed pipe is designed on the principle of the vacuum pump in the area of the opening of the ventilation pipe. There is a sub-pressure in the ventilation pipe, which makes it possible to extract the steam or gas. After mixing in the area of the opening, the steam condenses or is distributed with the gas in the primary coolant. (orig.) [de

  11. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  12. 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

  13. Cooling facility for reactor container

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro

    1996-05-31

    A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

  14. Cooling facility for reactor container

    International Nuclear Information System (INIS)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro.

    1996-01-01

    A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

  15. Cooling System Design Options for a Fusion Reactor

    Science.gov (United States)

    Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

    1997-06-01

    The objective of a fusion power reactor is to produce electricity safely and reliably. Accordingly, the design, objective of the heat transport system is to optimize power production, safety, and reliability. Such an optimization process, however, is constrained by many factors, including, among others: public safety, worker safety, steam cycle efficiency, reliability, and cost. As these factors impose conflicting requirements, there is a need to find an optimum design solution, i.e., one that satisfies all requirements, but not necessarily each requirement optimally. The SEAFP reactor study developed helium-cooled and water-cooled models for assessment purposes. Among other things, the current study demonstrates that neither model offers an optimum solution. Helium cooling offers a high steam cycle efficiency but poor reliability for the cooling of high heat flux components (divertor and first wall). Alternatively, water cooling offers a low steam cycle efficiency, but reasonable reliability for the cooling of such components. It is concluded that an optimum solution includes helium cooling of low heat flux components and water cooling of high heat flux components. Relative to the SEAFP helium model, this hybrid system enhances safety and reliability, while retaining the high steam cycle efficiency of that model.

  16. 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

  17. 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

  18. Cooling device for reactor suppression pool

    International Nuclear Information System (INIS)

    Togasaki, Susumu; Kato, Kiyoshi.

    1994-01-01

    In a cooling device of a reactor suppression pool, when a temperature of pool water is abnormally increased and a heat absorbing portion is heated by, for example, occurrence of an accident, coolants are sent to the outside of the reactor container to actuates a thermally operating portion by the heat energy of coolants and drive heat exchanging fluids of a secondary cooling system. If the heat exchanging fluids are sent to a cooling portion, the coolants are cooled and returned to the heat absorbing portion of the suppression pool water. If the heat absorbing portion is heat pipes, the coolants are evaporated by heat absorbed from the suppression pool water, steams are sent to the thermally operating portion, then coolants are liquefied and caused to return to the heat absorbing portion. If the thermal operation portion is a gas turbine, the gas turbine is operated by the coolants, and it is converted to a rotational force to drive heat exchanging fluids by pumps. By constituting the cooling portion with a condensator, the coolants are condensed and liquefied and returned to the heat absorbing portion of the suppression pool water. (N.H.)

  19. ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor

    International Nuclear Information System (INIS)

    Conklin, J.C.

    2001-01-01

    1 - Description of program or function: ORTURB was written specifically to calculate the dynamic behavior of the Fort St. Vrain (FSV) High- Temperature Gas-Cooled Reactor (HTGR) steam turbines. The program is divided into three main parts: the driver subroutine; turbine subroutines to calculate the pressure-flow balance of the high-, intermediate-, and low-pressure turbines; and feedwater heater subroutines. 2 - Method of solution: The program uses a relationship derived for ideal gas flow in an iterative fashion that minimizes computational time to determine the pressure and flow in the FSV steam turbines as a function of plant transient operating conditions. An important computer modeling characteristic, unique to FSV, is that the high-pressure turbine exhaust steam is used to drive the reactor core coolant circulators prior to entering the reheater. A feedwater heater dynamic simulation model utilizing seven state variables for each of the five heaters is included in the ORTURB computer simulation of the regenerative Rankine cycle steam turbines. The seven temperature differential equations are solved at each time- step using a matrix exponential method. 3 - Restrictions on the complexity of the problem: The turbine shaft is assumed to rotate at a constant (rated) speed of 3600 rpm. Energy and mass storage of steam in the high-, intermediate-, and low-pressure turbines is assumed to be negligible. These limitations exclude the use of ORTURB during a turbine transient such as startup from zero power or very low turbine flows

  20. Material choices for the commercial fast reactor steam generators

    International Nuclear Information System (INIS)

    Willby, C.; Walters, J.

    1978-01-01

    Experience with fast reactor steam generators has shown them to be critical components in achieving a high availability. This paper presents the designers views on the use of ferritic materials for steam generators and describes the proposed design of the steam generators for the Commercial Fast Reactor (CFR), prototype of which are to be inserted in the Prototype Fast Reactor at Dounreay. (author)

  1. Wet steam turbines for CANDU-Reactors

    International Nuclear Information System (INIS)

    Westmacott, C.H.L.

    1977-01-01

    The technical characteristics of 4 wet steam turbine aggregates used in the Pickering nuclear power station are reported on along with operational experience. So far, the general experience was positive. Furthermore, plans are mentioned to use this type of turbines in other CANDU reactors. (UA) [de

  2. Sodium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hammers, H.W.

    1982-01-01

    The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

  3. 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

  4. HTGR steam generator development

    International Nuclear Information System (INIS)

    Schuetzenduebel, W.G.; Hunt, P.S.; Weber, M.

    1976-01-01

    More than 40 gas-cooled reactor plants have produced in excess of 400 reactor years of operating experience which have proved a reasonably high rate of gas-cooled reactor steam generator availability. The steam generators used in these reactors include single U-tube and straight-tube steam generators as well as meander type and helically wound or involute tube steam generators. It appears that modern reactors are being equipped with helically wound steam generators of the once-through type as the end product of steam generator evolution in gas-cooled reactor plants. This paper provides a general overview of gas-cooled reactor steam generator evolution and operating experience and shows how design criteria and constraints, research and development, and experience data are factored into the design/development of modern helically wound tube steam generators for the present generation of gas-cooled reactors

  5. Emergency reactor container cooling facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Matsumoto, Tomoyuki.

    1992-01-01

    The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

  6. Emergency cooling apparatus for reactor

    International Nuclear Information System (INIS)

    Sakaguchi, S.

    1975-01-01

    A nuclear reactor is described which has the core surrounded by coolant and an inert cover gas all sealed within a container, an emergency cooling apparatus employing a detector that will detect cover gas or coolant, particularly liquid sodium, leaking from the container of the reactor, to release a heat exchange material that is inert to the coolant, which heat exchange material is cooled during operation of the reactor. The heat exchange material may be liquid niitrogen or a combination of spheres and liquid nitrogen, for example, and is introduced so as to contact the coolant that has leaked from the container quickly so as to rapidly cool the coolant to prevent or extinguish combustion. (Official Gazette)

  7. Process and device for accelerating condensation of the steam produced during an accident from the pressure vessel of a water cooled nuclear reactor

    International Nuclear Information System (INIS)

    Schnitker, W.

    1980-01-01

    In case of an accident, the steam from the PWR is taken away via lances under the water surface of the condensation area. In order to accelerate condensation, water is added via pipes projecting sideways into the lances. The kinetic energy of the steam carries the water over and produces a fog. (DG) [de

  8. Process and device for accelerating condensation of the steam produced during an accident from the pressure vessel of a water cooled nuclear reactor

    International Nuclear Information System (INIS)

    Schnitker, W.

    1981-01-01

    In case of an accident, the steam from the PWR is taken away via lances under the water surface of the condensation area. In order to accelerate condensation, water is added via pipes projecting sideways into the lances. The kinetic energy of the steam carries the water over and produces a fog. (orig./PW)

  9. 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

  10. Steam generator for nuclear reactors

    International Nuclear Information System (INIS)

    Byerley, W.M.; Bennett, R.R.

    1978-01-01

    In the steam generator, the primary medium is led through a U-shaped tube bundle heating up a secondary medium (feedwater) which flows around the tube bundle via a preheating chamber. In order to optimize heat transfer inside the preheating chamber, the feedwater is separated into a counterflow and a parallel flow with regard to the primary medium by means of partitioning walls and deflectors. The ratio is 70/30%. This way, boiling in the preheater is avoided, i.e. the high LMTD (logaritmic mean temperature difference) is fully utilized. (DG) [de

  11. Dry cooling tower operating experience in the LOFT reactor

    International Nuclear Information System (INIS)

    Hunter, J.A.

    1980-01-01

    A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

  12. Emergency cooling system for nuclear reactors

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    Upon the occasion of loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. Since the guide thimbles are mounted at predetermined positions relative to heat generating fuel elements in the fuel assemblies, holes bored at selected locations in the guide thimble walls, sprays the coolant against the reactor fuel elements which continue to dissipate heat but at a reduced level. The cooling water evaporates upon contacting the fuel rods thereby removing the maximum amount of heat (970 BTU per pound of water) and after heat absorption will leave the reactor in the form of steam through the break which is the cause of the accident to help assure immediate core cooldown

  13. Rod Bundle Heat Transfer: Steady-State Steam Cooling Experiments

    International Nuclear Information System (INIS)

    Spring, J.P.; McLaughlin, D.M.

    2006-01-01

    Through the joint efforts of the Pennsylvania State University and the United States Nuclear Regulatory Commission, an experimental rod bundle heat transfer (RBHT) facility was designed and built. The rod bundle consists of a 7 x 7 square pitch array with spacer grids and geometry similar to that found in a modern pressurized water reactor. From this facility, a series of steady-state steam cooling experiments were performed. The bundle inlet Reynolds number was varied from 1 400 to 30 000 over a pressure range from 1.36 to 4 bars (20 to 60 psia). The bundle inlet steam temperature was controlled to be at saturation for the specified pressure and the fluid exit temperature exceeded 550 deg. C in the highest power tests. One important quantity of interest is the local convective heat transfer coefficient defined in terms of the local bulk mean temperature of the flow, local wall temperature, and heat flux. Steam temperatures were measured at the center of selected subchannels along the length of the bundle by traversing miniaturized thermocouples. Using an analogy between momentum and energy transport, a method was developed for relating the local subchannel centerline temperature measurement to the local bulk mean temperature. Wall temperatures were measured using internal thermocouples strategically placed along the length of each rod and the local wall heat flux was obtained from an inverse conduction program. The local heat transfer coefficient was calculated from the data at each rod thermocouple location. The local heat transfer coefficients calculated for locations where the flow was fully developed were compared against several published correlations. The Weisman and El-Genk correlations were found to agree best with the RBHT steam cooling data, especially over the range of turbulent Reynolds numbers. The effect of spacer grids on the heat transfer enhancement was also determined from instrumentation placed downstream of the spacer grid locations. The local

  14. Reactor cooling apparatus

    International Nuclear Information System (INIS)

    Ogura, Kenji.

    1983-01-01

    Purpose: To increase natural convection flowrate in the reactor core upon interruption of a recycling pump by remarkably decreasing the flow resistance. Constitution: By-pass lines are disposed to a recycling pump in a primary coolant system and a second recycling pump in a secondary coolant system respectively, and a check valve and an isolation valve are attached to each of them. Each of the isolation valves is closed during normal operation and automatically opened when the number of rotation for each of the recycling pumps goes lower than a predetermined value. This can significantly decrease the flow resistance in the primary and secondary coolant systems upon interruption of the recycling pumps due to the entire loss of AC power source or the like to thereby increase the natural convection flowrate in the reactor core. (Sekiya, K.)

  15. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Leigh, K.M.

    1980-01-01

    A liquid metal cooled nuclear reactor is described, wherein coolant is arranged to be flowed upwardly through a fuel assembly and having one or more baffles located above the coolant exit of the fuel assembly, the baffles being arranged so as to convert the upwardly directed motion of liquid metal coolant leaving the fuel assembly into a substantially horizontal motion. (author)

  16. Cooling system upon reactor isolation

    International Nuclear Information System (INIS)

    Yamamoto, Kohei; Oda, Shingo; Miura, Satoshi

    1992-01-01

    A water level indicator for detecting the upper limit value for a range of using a suppression pool and a thermometer for detecting the temperature of water at the cooling water inlet of an auxiliary device are disposed. When a detection signal is intaken and the water level in the suppression pool reach the upper limit value for the range of use, a secondary flow rate control value is opened and a primary flow rate control valve is closed. When the temperature of the water at the cooling water inlet of the auxiliary device reaches the upper limit value, the primary and the secondary flow rate control valves are opened. During a stand-by state, the first flow rate control valve is set open and the secondary flow rate control valve is set closed respectively. After reactor isolation, if a reactor water low level signal is received, an RCIC pump is actuated and cooling water is sent automatically under pressure from a condensate storage tank to the reactor and the auxiliary device requiring coolants by way of the primary flow rate control valve. Rated flow rate is ensured in the reactor and cooling water of an appropriate temperature can be supplied to the auxiliary device. (N.H.)

  17. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  18. 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

  19. Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1983-01-01

    This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

  20. Helium cooling of fusion reactors

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

    1994-01-01

    On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

  1. Liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Filonov, V.S.; Zaitsev, B.I.; Artemiev, L.N.; Rakhimov, V.V.

    1976-01-01

    A liquid-metal-cooled reactor is described comprising two rotatable plugs, one of them, having at least one hole, being arranged internally of the other, a recharging mechanism with a guide tube adapted to be moved through the hole of the first plug by means of a drive, and a device for detecting stacks with leaky fuel elements, the recharging mechanism tube serving as a sampler

  2. Emergency reactor core cooling facility

    International Nuclear Information System (INIS)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka.

    1996-01-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  3. Emergency reactor core cooling facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

    1996-11-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  4. Emergency cooling device for reactors

    International Nuclear Information System (INIS)

    Inoue, Hisamichi; Naito, Masanori; Sato, Chikara; Chino, Koichi.

    1975-01-01

    Object: To pour high pressure cooling water into a core, when coolant is lost in a boiling water reactor, thereby restraining the rise of fuel cladding. Structure: A control rod guiding pipe, which is moved up and down by a control rod, is mounted on the bottom of a pressure vessel, the control rod guiding pipe being communicated with a high pressure cooling water tank positioned externally of the pressure vessel, and a differential in pressure between the pressure vessel and the aforesaid tank is detected when trouble of coolant loss occurs, and the high pressure cooling water within the tank is poured into the core through the control rod guiding pipe to restrain the rise of fuel cladding. (Kamimura, M.)

  5. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

    A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

  6. Feasibility and application on steam injector for next-generation reactor

    International Nuclear Information System (INIS)

    Narabayashi, Tadashi; Ishiyama, Takenori; Miyano, Hiroshi; Nei, Hiromichi; Shioiri, Akio

    1991-01-01

    A feasibility study has been conducted on steam injector for a next generation reactor. The steam injector is a simple, compact passive device for water injection, such as Passive Core Injection System (PCIS) of Passive Containment Cooling System (PCCS), because of easy start-up without an AC power. An analysis model for a steam injector characteristics has been developed, and investigated with a visualized fundamental test for a two-stage Steam Injector System (SIS) for PCIS and a one-stage low pressure SIS for PCCS. The test results showed good agreement with the analysis results. The analysis and the test results showed the SIS could work over a very wide range of the steam pressure, and is applicable for PCIS or PCCS in the next generation reactors. (author)

  7. Steam generator of FBR type reactor

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1992-01-01

    Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

  8. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Guidez, Joel; Jarriand, Paul.

    1975-01-01

    The invention concerns a fast neutron nuclear reactor cooled by a liquid metal driven through by a primary pump of the vertical drive shaft type fitted at its lower end with a blade wheel. To each pump is associated an exchanger, annular in shape, fitted with a central bore through which passes the vertical drive shaft of the pump, its wheel being mounted under the exchanger. A collector placed under the wheel comprises an open upward suction bell for the liquid metal. A hydrostatic bearing is located above the wheel to guide the drive shaft and a non detachable diffuser into which at least one delivery pipe gives, envelopes the wheel [fr

  9. Investigation of vessel exterior air cooling for a HLMC reactor

    International Nuclear Information System (INIS)

    Sienicki, J. J.; Spencer, B. W.

    2000-01-01

    The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink

  10. Emergency core cooling system for LMFBR type reactors

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Fukutomi, Shigeki.

    1980-01-01

    Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

  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. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  13. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    A concrete containment vault for a liquid metal cooled nuclear reactor is described which is lined with thermal insulation to protect the vault against heat radiated from the reactor during normal operation of the reactor but whose efficiency of heat insulation is reduced in an emergency to enable excessive heat from the reactor to be dissipated through the vault. (UK)

  14. Passive cooling applications for nuclear power plants using pulsating steam-water heat pipes

    International Nuclear Information System (INIS)

    Aparna, J.; Chandraker, D.K.

    2015-01-01

    Gen IV reactors incorporate passive principles in their system design as an important safety philosophy. Passive safety systems use inherent physical phenomena for delivering the desired safe action without any external inputs or intrusion. The accidents in Fukushima have renewed the focus on passive self-manageable systems capable of unattended operation, for long hours even in extended station blackout (SBO) and severe accident conditions. Generally, advanced reactors use water or atmospheric air as their ultimate heat sink and employ passive principles in design for enhanced safety. This paper would be discussing the experimental results on pulsating steam water heat-pipe devices and their applications in passive cooling. (author)

  15. Steam generator

    International Nuclear Information System (INIS)

    Fenet, J.-C.

    1980-01-01

    Steam generator particularly intended for use in the coolant system of a pressurized water reactor for vaporizing a secondary liquid, generally water, by the primary cooling liquid of the reactor and comprising special arrangements for drying the steam before it leaves the generator [fr

  16. Environmental, financial and energy feasibility of the electrolysis of the water steam of a generation IV reactor cooling system during the moments of low consumption ef eletrical energy: challenges and perspectives

    International Nuclear Information System (INIS)

    Stefanelli, Eduardo J.; Vargas, Miltom; Garcia, Pedro L.; Seo, Emilia S.M.; Oliveira, Wagner de S.

    2009-01-01

    Our civilization is in an inflection moment. Our current decisions will lead us to drastic changes in the planet climate or to cleaner and more sustainable energy generation models. In a certain moment in the evolution of our society, we have privileged the productivity instead of the reasonable use of the planet's resources. This option has been leading us to a situation in which these resources are seriously jeopardized. New models of energy generation and use should be discussed and adopted in order to reverse this process. The electric-power consumption is not constant through time and it must be generated at the moment it is going to be used. There are moments of great idleness in the electric-power generation system, counterbalanced by high demand moments. This characteristic has induced us to the construction of a model of great generation capacity that remains without use most of the time, producing huge financial and environmental impacts. In this article, we discuss the environmental, financial and energy viability of using the idle capacity of the electric-power system to, through water steam electrolysis, produce hydrogen, which would be reconverted into electric power in peak moments by a fuel cell. In this study, we aim at investigating the viability of associating a SOFC (Solid Oxide Fuel Cell), acting as an electrolysis bow, to a generation IV reactor, in order to produce hydrogen from superheated water steam in the cooling of the reactor, which will be converted into electric power via SOFC (Solid Oxide Fuel Cell) in peak moments. The method used in this investigation was to study the electric charge variation consumed in a day, randomly selected in relation to the hour of the day, to launch a curve into a diagram 'Demand x hour of the day', to establish the peak moments, the minimum moments, and the average consumption, and, based on these data and geometrically, predict the viability of using the energetic potential of the moments in which the

  17. Environmental, financial and energy feasibility of the electrolysis of the water steam of a generation IV reactor cooling system during the moments of low consumption ef eletrical energy: challenges and perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Stefanelli, Eduardo J.; Vargas, Miltom; Garcia, Pedro L.; Seo, Emilia S.M.; Oliveira, Wagner de S. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], e-mail: ipen@stefanelli.eng.br

    2009-07-01

    Our civilization is in an inflection moment. Our current decisions will lead us to drastic changes in the planet climate or to cleaner and more sustainable energy generation models. In a certain moment in the evolution of our society, we have privileged the productivity instead of the reasonable use of the planet's resources. This option has been leading us to a situation in which these resources are seriously jeopardized. New models of energy generation and use should be discussed and adopted in order to reverse this process. The electric-power consumption is not constant through time and it must be generated at the moment it is going to be used. There are moments of great idleness in the electric-power generation system, counterbalanced by high demand moments. This characteristic has induced us to the construction of a model of great generation capacity that remains without use most of the time, producing huge financial and environmental impacts. In this article, we discuss the environmental, financial and energy viability of using the idle capacity of the electric-power system to, through water steam electrolysis, produce hydrogen, which would be reconverted into electric power in peak moments by a fuel cell. In this study, we aim at investigating the viability of associating a SOFC (Solid Oxide Fuel Cell), acting as an electrolysis bow, to a generation IV reactor, in order to produce hydrogen from superheated water steam in the cooling of the reactor, which will be converted into electric power via SOFC (Solid Oxide Fuel Cell) in peak moments. The method used in this investigation was to study the electric charge variation consumed in a day, randomly selected in relation to the hour of the day, to launch a curve into a diagram 'Demand x hour of the day', to establish the peak moments, the minimum moments, and the average consumption, and, based on these data and geometrically, predict the viability of using the energetic potential of the moments

  18. 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

  19. Application of a steam injector for passive emergency core cooling during a station blackout

    International Nuclear Information System (INIS)

    Heinze, D.; Behnke, L.; Schulenberg, T.

    2012-01-01

    One of the basic protection targets of reactor safety is the safe heat removal during normal operation but also following shut-down. Since the reactor accident in Fukushima an optimization of the plant robustness in case of beyond-design accident is performed. Special attention is given to the increase of time available for starting appropriate measures for emergency core cooling in case of a station blackout. The state-of the art in engineering and research is presented. Investigations on the applicability of a steam injector for passive emergency core cooling during a station blackout in BWR-type reactors have progressed, experiments on dynamic behavior of the injector are described. A precise design with respect to the thermal hydraulic boundary conditions has been performed.

  20. Advanced steam cycles for light water reactors. Final report

    International Nuclear Information System (INIS)

    Mitchell, R.C.

    1975-07-01

    An appraisal of the potential of adding superheat to improve the overall LWR plant cycle performance is presented. The study assesses the economic and technical problems associated with the addition of approximately 500 0 F of superheat to raise the steam temperature to 1000 0 F. The practicality of adding either nuclear or fossil superheat to LWR's is reviewed. The General Electric Company Boiling Water Reactor (BWR) model 238-732 (BWR/6) is chosen as the LWR starting point for this evaluation. The steam conditions of BWR/6 are representative of LWR's. The results of the fossil superheat portion of the evaluation are considered directly applicable to all LWR's. In spite of the potential of a nuclear superheater to provide a substantial boost to the LWR cycle efficiency, nuclear superheat offers little promise of development at this time. There are difficult technical problems to resolve in the areas of superheat fuel design and emergency core cooling. The absence of a developed high integrity, high temperature fuel for operation in the steam/water environment is fundamental to this conclusion. Fossil superheat offers the potential opportunity to utilize fossil fuel supplies more efficiently than in any other mode of central station power generation presently available. Fossil superheat topping cycles evaluated included atmospheric fluidized beds (AFB), pressurized fluidized beds, pressurized furnaces, conventional furnaces, and combined gas/steam turbine cycles. The use of an AFB is proposed as the preferred superheat furnace. Fossil superheat provides a cycle efficiency improvement for the LWR of two percentage points, reduces heat rejection by 15 percent per kWe generated, increases plant electrical output by 54 percent, and burns coal with an incremental net efficiency of approximately 40 percent. This compares with a net efficiency of 36--37 percent which might be achieved with an all-fluidized bed fossil superheat plant design

  1. The steam generating heavy water reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    A review is presented on the evolution of the SGHWR concept by the United Kingdom Atomic Energy Authority and the production of early commercial designs, together with later development by the Design and Construction Companies. This is followed by a description of the current commercial design. Possible future developments are suggested. The many advantageous features of the concept are mentioned with a view to supporting optimism for the future of the system. Headings include the following: safety criteria and risk assessment; emergency core cooling system design and development; protective systems; reactor coolant system; reactivity control; off-load refuelling; pressure containment; 'fence' header coolant circuit design; feed water injection; continuous spray cooling; low pressure cooling systems for residual heat removal during refuelling; high pressure cooling system for guaranteed feed water supply; auxiliary systems; structural materials; calandria and neutron shields; fuel element development; alternative loop circuit design; future developments (use of hydraulic diodes to provide a substantial reverse flow resistance by the generation of a vortex; multi-drum and multi-pump schemes; refuelling alternatives; coolant circuit inversion; use of superheat channels). (U.K.)

  2. 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

  3. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    Directory of Open Access Journals (Sweden)

    Fic Adam

    2015-03-01

    Full Text Available Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle, which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle. The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

  4. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

    A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

  5. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  6. Graphite-water steam-generating reactor in the USSR

    Energy Technology Data Exchange (ETDEWEB)

    Dollezhal, N A [AN SSSR, Moscow

    1981-10-01

    One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

  7. 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.

  8. 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)

  9. Design and analysis on super-critical water cooled power reactors

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki

    2005-01-01

    The Super-Critical Water Cooled Power Reactors (SCPR) is cooled by 25 MPa supercritical water of 280degC at reactor inlet and greater than 500degC at reactor outlet and directly connected with turbine/generators with high energy conversion efficiency. This corresponds to the deletion of recirculation system and steam-water separation system of BWR type reactors or of pressurizer and steam generator of PWR type reactors. In addition to the design study of the university of Tokyo, technology development of the SCPR for practical use has started under the collaboration of industry and academia since 2000. Mockup single tube and bundle tests for heat transfer/fluid flow characteristics of the design have been conducted with 3D heat transfer analysis. Materials compatible with coolant conditions for fuel cans and reactor internals are also assessed. Overall evaluation of the reactor concept is under way. (T. Tanaka)

  10. Cooling device upon reactor isolation

    International Nuclear Information System (INIS)

    Otsu, Tatsuya

    1995-01-01

    A vacuum breaking valve is disposed to a sucking pipeline of vacuum pumps. A sucking port of the breaking valve is connected with an exhaustion side of a relief valve of a liquid nitrogen-filled tank by way of communication pipes. When a cooling device is operated upon reactor isolation and the vacuum pumps are operated, a three directional electromagnetic valve is operated, and nitrogen discharged out of the exhaustion port of the relief valve of the liquid nitrogen-filled tank is sent to a nitrogen releasing port on the suction side of the vacuum breaking valve by way of the communication pipes and released to atmosphere. When the pressure in the vacuum tank is excessively lowered in this state and the vacuum breaking valve is opened, nitrogen flows from the nitrogen discharge port into the vacuum tank through the breaking valve, and are sent to a pressure suppression chamber by the vacuum pumps. Since a great amount of nitrogen is sent to the pressure suppression chamber, and the inflow of the air is reduced, increase of oxygen concentration in the pressure suppression chamber can be suppressed. (I.N.)

  11. Study on steam separation in steam generators of a NPP with the WWER-440 reactors

    International Nuclear Information System (INIS)

    Dmitriev, A.I.; Kolzov, Yu.V.; Titov, V.F.; Dubrovin, A.V.; Ilyushin, V.F.; Volkov, A.P.

    1977-01-01

    The separation characteristics as well as the actual level position in steam generators with and without a submerged holy sheet have been determined at a WWER-440 reactor nuclear power plant. It has been shown, that without changing the design of steam generators their load at the WWER-440 reactor nuclear power plant can be increased by about 10%. In this case the vapour humidity does not exceed the permissible value equal to 0.25%. The submerged holy sheet considerably decreases load irregularity and swelling of the water-steam mixture layer

  12. 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.

  13. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  14. Clinch river breeder reactor plant steam generator water quality

    International Nuclear Information System (INIS)

    Van Hoesen, D.; Lowe, P.A.

    1975-01-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

  15. Clinch river breeder reactor plant steam generator water quality

    Energy Technology Data Exchange (ETDEWEB)

    Van Hoesen, D; Lowe, P A

    1975-07-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

  16. Steam line break analysis in CAREM-25 reactor

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo O.; Vertullo, Alicia; Schlamp, Miguel A.; Garcia, Alicia E.

    2000-01-01

    The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model. The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator. As a consequence and due to reactor features the core power is also increased. As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided. Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System. In all the sequences the DNBR and CPR remain above the minimum safety values established by design. Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised. (author)

  17. Steam Line Break Analysis in CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Zanocco, Pablo; Gimenez, Marcelo; Vertullo, Alicia; Garcia, A; Schlamp, Miguel

    2000-01-01

    The main objective of this report is to analyze the reactor response during a steam line break postulated accident with RELAP5, a plant code using a separated flow model.The steam line break caused a rapid blowdown of the secondary coolant increasing the heat removal in the steam generator.As a consequence and due to reactor features the core power is also increased.As maximum removed power in the secondary side is highly dependant on the total water volume evaporated during the accident a detailed model of feed water and outlet steam pipes is provided.Different cases are analyzed with and without feedwater system and considering the fail or success of the First Shutdown System.In all the sequences the DNBR and CPR remain above the minimum safety values established by design.Further calculations concerning depressurization ramps and steam generator feed water pumps response during depressurization are advised

  18. Composite electric generator equipped with steam generator for heating reactor coolant

    International Nuclear Information System (INIS)

    Watabe, Masaharu; Soman, Yoshindo; Kawanishi, Kohei; Ota, Masato.

    1997-01-01

    The present invention concerns a composite electric generator having coolants, as a heating source, of a PWR type reactor or a thermonuclear reactor. An electric generator driving gas turbine is disposed, and a superheater using a high temperature exhaust gas of the gas turbine as a heating source is disposed, and main steams are superheated by the superheater to elevate the temperature at the inlet of the turbine. This can increase the electric generation capacity as well as increase the electric generation efficiency. In addition, since the humidity in the vicinity of the exit of the steam turbine is reduced, occurrence of loss and erosion can be suppressed. When cooling water of the thermonuclear reactor is used, the electric power generated by the electric generator driven by the gas turbine can be used upon start of the thermonuclear reactor, and it is not necessary to dispose a large scaled special power source in the vicinity, which is efficient. (N.H.)

  19. Limiting Factors for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Cheung, F.B.

    2005-01-01

    The method of external reactor vessel cooling (ERVC) that involves flooding of the reactor cavity during a severe accident has been considered a viable means for in-vessel retention (IVR). For high-power reactors, however, there are some limiting factors that might adversely affect the feasibility of using ERVC as a means for IVR. In this paper, the key limiting factors for ERVC have been identified and critically discussed. These factors include the choking limit for steam venting (CLSV) through the bottleneck of the vessel/insulation structure, the critical heat flux (CHF) for downward-facing boiling on the vessel outer surface, and the two-phase flow instabilities in the natural circulation loop within the flooded cavity. To enhance ERVC, it is necessary to eliminate or relax these limiting factors. Accordingly, methods to enhance ERVC and thus improve margins for IVR have been proposed and demonstrated, using the APR1400 as an example. The strategy is based on using two distinctly different methods to enhance ERVC. One involves the use of an enhanced vessel/insulation design to facilitate steam venting through the bottleneck of the annular channel. The other involves the use of an appropriate vessel coating to promote downward-facing boiling. It is found that the use of an enhanced vessel/insulation design with bottleneck enlargement could greatly facilitate the process of steam venting through the bottleneck region as well as streamline the resulting two-phase motions in the annular channel. By selecting a suitable enhanced vessel/insulation design, not only the CLSV but also the CHF limits could be significantly increased. In addition, the problem associated with two-phase flow instabilities and flow-induced mechanical vibration could be minimized. It is also found that the use of vessel coatings made of microporous metallic layers could greatly facilitate downward-facing boiling on the vessel outer surface. With vessel coatings, the local CHF limits at

  20. System design study of small lead-bismuth cooled reactor

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Hori, Toru; Konomura, Mamoru

    2003-07-01

    In phase II of the feasibility study of JNC, we will make a concept of a dispersion power source reactor with various requirements, such as economical competitiveness and safety. In the study of a small lead-bismuth cooled reactor, a concept whose features are long life core, inherent safety, natural convection of cooling system and steam generators in the reactor vessel has been designed since 2000. The investigations which have been done in 2002 are shown as follows; Safety analysis of UTOP considering uncertainty of reactivity. Possibility of reduction of number of control rods. Estimation of construction cost. Transient analyses of UTOP have been done in considering uncertainty of reactivity in order to show the inherent safety in the probabilistic method. And the inherent safety in UTOP is realized under the condition of considering uncertainty. Transient analyses of UTOP with various numbers of control rods have been done and it is suggested that there is possibility of reduction of the number of control rods considering accident managements. The method of cost estimation is a little modified. The cost of reactor vessel is estimated from that of medium sized lead-bismuth cooled reactor and the estimation of a purity control system is by coolant volume flow rate. The construction cost is estimated 850,000yen/kWe. (author)

  1. An improved water cooled nuclear reactor and pressuriser assembly

    International Nuclear Information System (INIS)

    Gardner, F.J.; Strong, R.

    1991-01-01

    A water cooled nuclear reactor is described which comprises a reactor core, a primary water coolant circuit and a pressuriser arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper and a lower chamber by a casing. The reactor core and primary water coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber. A plurality of spray pipes interconnect a steam space of the pressuriser with the downcomer of the primary water coolant circuit below a heat exchanger. A plurality of surge ports interconnect a water space of the pressuriser with the primary water coolant circuit. The surge ports have hydraulic diodes so that there is a low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and high flow resistance in the opposite direction. The spray pipes provide a desuperheating spray of cooled water into the pressuriser during positive volume surges of the primary water coolant. The pressuriser arrangement may also be applied to integral water cooled reactors with separate pressurisers and to dispersed pressurised water reactors. The surge ports also allow water to flow by gravity to the core in an emergency. (author)

  2. Steam plant for pressurized water reactors

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

  3. 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

  4. New safety experiments in decommissioned superheated steam reactor at Karlstein

    International Nuclear Information System (INIS)

    Koerting, K.

    1986-01-01

    This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

  5. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    Fabrega, S.

    1979-01-01

    The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

  6. 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)

  7. Hot steam header of a high temperature reactor as a benchmark problem

    International Nuclear Information System (INIS)

    Demierre, J.

    1990-01-01

    The International Atomic Energy Agency (IAEA) initiated a Coordinated Research Programme (CRP) on ''Design Codes for Gas-Cooled Reactor Components''. The specialists proposed to start with a benchmark design of a hot steam header in order to get a better understanding of the methods in the participating countries. The contribution of Switzerland carried out by Sulzer. The following report summarized the detailed calculations of dimensioning procedure and analysis. (author). 5 refs, 2 figs, 2 tabs

  8. 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

  9. Incoloy 800 stands up to radiation and corrosion in high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Incoloy 800 has been selected for heat exchangers in helium cooled nuclear reactor prototypes for exposure to 350 to 800 0 C helium and high temperature high purity water and steam. 304H stainless steel used in heat exchangers in original design cracked in the superheater area, bellows and tubing after static pressure tests but before exposure to steam. Residual stress, chlorides, and oxygen were deduced to have caused the failures

  10. Research of impact of kind resuperheat and structure of system regenerative feed water to thermodynamic efficiency of cycle with steam-coolant reactor

    Directory of Open Access Journals (Sweden)

    Maykova Svetlana

    2017-01-01

    Full Text Available The first key problems of modern nuclear reactors are inability of closed nuclear cycle, problems with spent nuclear fuel, poor effectiveness of nuclear fuel and heat-exchange equipment usage. Dealing with problems consists in usage of fast-neutron reactors with steam coolant. Scientific men analyzed neutron-physical processes in steam-cooled fast reactor and consulted that creation of the reactor is viable. In consequence of low steam activation a single-loop steam cycle may be create. The cycle is easy and fool-proof. Core thermomechanical equipment has mastered and has relatively low metal content. Results of calculation are showing that nuclear unit with steam-coolant fast neutron reactor is more efficient than widely used unit with reactor VVER. Usage of simple scheme with four regenerative feedwater heaters the absolute efficiency ratio is more than 43%.

  11. 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)

  12. Reactor physics studies in the steam flooded GCFR-Phase II critical assembly

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.

    1978-08-01

    A possible accident scenario in a Gas-Cooled Fast Reactor (GCFR) is the leakage of secondary steam into the core. Considerable analytical effort has gone into the study of the effects of such an accidental steam entry. The work described represents the first full scale experimental study of the steam-entry phenomenon in GCFRs. The reference GCFR model used for the study was the benchmark GCFR Phase II assembly, and polyethylene foam was used to provide a very homogeneous steam simulation. The reactivity worth of steam entry was measured for three different steam densities. In addition, a set of integral physics parameters were measured in the largest steam density (0.008 g/cm 3 ) configuration. The corresponding parameters were also measured in dry reference GCFR critical assembly for comparison. The experiments were analyzed using ENDF/B-IV data and two-dimensional diffusion theory methods. As in earlier GCFR critical experiments analysis, the Benoist method was used to treat the problem of neutron streaming

  13. On the reliability of steam generator performance at nuclear power plants with WWER type reactors

    International Nuclear Information System (INIS)

    Styrikovich, M.A.; Margulova, T.Kh.

    1974-01-01

    The problem of ensuring reliable operation of steam generators in a nuclear power plant with a water-cooled, water-moderated reactor (WWER) was studied. At a nuclear power plant with a vertical steam generator (specifically, a Westinghouse product) the steam generator tubes were found to have been penetrated. Shutdown was due to corrosion disintegration of the austenitic stainless steel, type 18/8, used as pipe material for the heater surface. The corrosion was the result of the action of chlorine ions concentrated in the moisture contained in the iron oxide films deposited in low parts of the tube bundle, directly at the tube plate. Blowing through did not ensure complete removal of the film, and in some cases the construction features of the steam generator made removal of the film practically impossible. Replacement of type 18/8 stainless steel by other construction material, e.g., Inconel, did not give good results. To ensure reliable operation of vertical steam generators in domestic practice, the generators are designed without a low tube plate (a variant diagram of the vertical steam generator of such construction for the water-cooled, water-moderated reactor 1000 is presented). When low tube plates are used the film deposition is intolerable. For organization of a non-film regime a complex treatment of the feed water is used, in which the amount of complexion is calculated from the stoichmetric ratios with the composition of the feed water. It is noted that, if 100% condensate purification is used with complexon processing of the feed water to the generator, we can calculate the surface of the steam-generator heater without considering the outer placement on the tubes. In this the cost of the steam generator and all the nuclear power plants with WWER type reactors is decreased even with installation of a 100% condensate purification. It is concluded that only simultaneous solution of construction and water-regime problems will ensure relaible operation of

  14. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  15. A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

  16. Helium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Longton, P.B.; Cowen, H.C.

    1975-01-01

    In helium cooled HTR's there is a by-pass circuit for cleaning purposes in addition to the main cooling circuit. This is to remove such impurities as hydrogen, methane, carbon monoxide and water from the coolant. In this system, part of the coolant successively flows first through an oxidation bed of copper oxide and an absorption bed of silica gel, then through activated charcoal or a molecular sieve. The hydrogen and carbon monoxide impurities are absorbed and the dry gas is returned to the main cooling circuit. To lower the hydrogen/water ratio without increasing the hydrogen fraction in the main cooling circuit, some of the hydrogen fraction converted into water is added to the cooling circuit. This is done, inter alia, by bypassing the water produced in the oxidation bed before it enters the absorption bed. The rest of the by-pass circuit, however, also includes an absorption bed with a molecular sieve. This absorbs the oxidized carbon monoxide fraction. In this way, such side effects as the formation of additional methane, carburization of the materials of the by-pass circuit or loss of graphite are avoided. (DG/RF) [de

  17. 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

  18. 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

  19. Design and selection of materials for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.

    2011-01-01

    Sodium cooled fast reactors are currently in operation, under construction or under design by a number of countries. The design of sodium cooled fast reactor is covered by French RCC - MR code and ASME code NH. The codes cover rules as regards to materials, design and construction. These codes do not cover the effect of irradiation and environment. Elevated temperature design criteria in nuclear codes are much stringent in comparison to non nuclear codes. Sodium corrosion is not an issue in selection of materials provided oxygen impurity in sodium is controlled for which excellent reactor operating experience is available. Austenitic stainless steels have remained the choice for the permanent structures of primary sodium system. Stabilized austenitic stainless steel are rejected because of poor operating experience and non inclusion in the design codes. Route for improved creep behaviour lies in compositional modifications in 316 class steel. However, the weldability needs to be ensured. For cold leg component is non creep regime, SS 304 class steel is favoured from overall economics. Enhanced fuel burn up can be realized by the use of 9-12%Cr 1%Mo class steel for the wrapper of MOX fuel design, and cladding and wrapper for metal fuel reactors. Minor compositional modifications of 20% cold worked 15Cr-15Ni class austenitic stainless steel will be a strong candidate for the cladding of MOX fuel design in the short term. Long term objective for the cladding will be to develop oxide dispersion strengthened steel. 9%Cr 1%Mo class steel (Gr 91) is an ideal choice for integrated once through sodium heated steam generators. One needs to incorporate operating experience from reactors and thermal power stations, industrial capability and R and D feedback in preparing the technical specifications for procurement of wrought products and welding consumables to ensure reliable operation of the components and systems over the design life. The paper highlights the design approach

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  1. 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.)

  2. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

    1989-06-01

    This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

  3. Modelling aerosol behavior in reactor cooling systems

    International Nuclear Information System (INIS)

    McDonald, B.H.

    1990-01-01

    This paper presents an overview of some of the areas of concern in using computer codes to model fission-product aerosol behavior in the reactor cooling system (RCS) of a water-cooled nuclear reactor during a loss-of-coolant accident. The basic physical processes that require modelling include: fission product release and aerosol formation in the reactor core, aerosol transport and deposition in the reactor core and throughout the rest of the RCS, and the interaction between aerosol transport processes and the thermalhydraulics. In addition to these basic physical processes, chemical reactions can have a large influence on the nature of the aerosol and its behavior in the RCS. The focus is on the physics and the implications of numerical methods used in the computer codes to model aerosol behavior in the RCS

  4. Steam generator for pressurized-water reactors

    International Nuclear Information System (INIS)

    Michel, E.

    1971-01-01

    In the steam generator for a PWR the central fall space of a U-tube bundel heat exchanger is used as a preliminary cyclon separator. The steam escaping upwards, which is largely free of water, can flow through the residual heating surface, i.e. the U-tube turns. In this way substantial drying and less superheating by the heat still added becomes possible. In its upper part the central fall space for the water separated in the preliminary separator, enclosed by a cylindrical guide wall and the U-tube bundle, is provided with tangential inlet slots. Through these, the water-steam mixture steams out of the section of the vertical legs of the U-tube bundle into the fall space. Above the inlet slots the rising space is closed by means of a turn-round plate. At the lower end of the guide wall outlet, slots are provided for the water flowing downwards and radially outwards into the unfilled space. (DG/PB) [de

  5. Processes influencing cooling of reactor effluents

    International Nuclear Information System (INIS)

    Magoulas, V.E.; Murphy, C.E. Jr.

    1982-01-01

    Discharge of heated reactor cooling water from SRP reactors to the Savannah River is through sections of stream channels into the Savannah River Swamp and from the swamp into the river. Significant cooling of the reactor effluents takes place in both the streams and swamp. The majority of the cooling is through processes taking place at the surface of the water. The major means of heat dissipation are convective transfer of heat to the air, latent heat transfer through evaporation and radiative transfer of infrared radiation. A model was developed which incorporates the effects of these processes on stream and swamp cooling of reactor effluents. The model was used to simulate the effect of modifications in the stream environment on the temperature of water flowing into the river. Environmental effects simulated were the effect of changing radiant heat load, the effect of changes in tree canopy density in the swamp, the effect of total removal of trees from the swamp, and the effect of diverting the heated water from L reactor from Steel Creek to Pen Branch. 6 references, 7 figures

  6. Steam turbine chemistry in light water reactor plants

    International Nuclear Information System (INIS)

    Svoboda, Robert; Haertel, Klaus

    2008-01-01

    Steam turbines in boiling water reactor (BWR) and pressurized water reactor (PWR) power plants of various manufacturers have been affected by corrosion fatigue and stress corrosion cracking. Steam chemistry has not been a prime focus for related research because the water in nuclear steam generating systems is considered to be of high purity. Steam turbine chemistry however addresses more the problems encountered in fossil fired power plants on all volatile treatment, where corrosive environments can be formed in zones where wet steam is re-evaporated and dries out, or in the phase transition zone, where superheated steam starts to condense in the low-pressure (LP) turbine. In BWR plants the situation is aggravated by the fact that no alkalizing agents are used in the cycle, thus making any anionic impurity immediately acidic. This is illustrated by case studies of pitting corrosion of a 12 % Cr steel gland seal and of flow-oriented corrosion attack on LP turbine blades in the phase transition zone. In PWR plants, volatile alkalizing agents are used that provide some buffering of acidic impurities, but they also produce anionic decomposition products. (orig.)

  7. IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Presentations

    International Nuclear Information System (INIS)

    2011-01-01

    The fast reactor, which can generate electricity and breed additional fissile material for future fuel stocks, is a resource that will be needed when economic uranium supplies for the thermal reactors diminish. Further, the fast-fission fuel cycle in which material is recycled (a basic requirement to meet sustainability criteria) offers the flexibility needed to contribute decisively towards solving the problem of growing “spent” fuel inventories by greatly reducing the volume, the heat load and the radiotoxic inventory of high-level wastes that must be disposed of in long-term geological repositories. This is a waste management option that will play an increasingly important role in the future, and help to ensure that nuclear energy remains a sustainable long-term option in the world’s overall energy mix. In recognition of the fast reactor’s importance for the sustainability of the nuclear option, currently there is worldwide renewed interest in fast reactor technology development, as indicated, e.g., by the outcome of the Generation IV International Forum (GIF) technology review, which concluded with 3 out of 6 innovative systems to be fast reactors (gas cooled fast reactor, sodium cooled fast reactor, and heavy liquid metal cooled fast reactor), plus a potential fast core for a 4th concept, the super-critical water reactor. Currently, fast reactor construction projects are ongoing in India (PFBR) and Russian Federation (BN-800), whilst in China the first experimental fast reactor (CEFR) is in the commissioning phase. Fast reactor programs are also carried out in Europe (in particular in France), Japan, Republic of Korea and the USA. The most important challenges for fast reactors are in the areas of cost competitiveness with respect to LWRs and other energy sources, enhanced safety, non-proliferation, and public acceptance. With the exception of this latter, these translate into technology development challenges, i.e. the development of advanced reactor

  8. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

  9. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Scott, D.

    1981-01-01

    An improved method of constructing the diagrid used to support fuel assemblies of liquid metal fast breeder reactors, is described. The functions of fuel assembly support and coolant plenum are performed by discrete components of the diagrid each of which can serve the function of the other in the event of failure of one of the components. (U.K.)

  10. 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

  11. 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

  12. Steam generator for PWR type reactor

    International Nuclear Information System (INIS)

    Baba, Iwao; Hiyama, Nobuyuki.

    1994-01-01

    A steam generator of the present invention comprises a primary coolant chamber having primary coolants circulating therein, a secondary coolants chamber having secondary coolants and steams circulating therein, which are isolated from each other by a partition wall, and heat pipes disposed being passed through the partition wall. The heat pipes are disposed having an evaporation portion in the primary coolants chamber, a condensation portion in the secondary coolants chamber, and an intermediate heat insulating portion in the partition wall. Since the primary coolants containing radioactivity and the secondary coolants not containing radioactivity does not transfer heat directly by a heat transfer wall, a leakage accident of radioactivity to the secondary coolants can be prevented. Moreover, since the heat pipes are used, a great amount of heat can be transferred by a slight temperature difference by using steams of the heat transfer medium itself, latent heat due to coagulation, and capillary phenomenon. Since neither transferring power nor pumps are required, heat of the primary coolants can effectively be transferred to the secondary coolants. (N.H.)

  13. 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

  14. 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

  15. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  16. The effect of cooling conditions on convective heat transfer and flow in a steam-cooled ribbed duct

    International Nuclear Information System (INIS)

    Shui, Linqi; Gao, Jianmin; Shi, Xiaojun; Liu, Jiazeng; Xu, Liang

    2014-01-01

    This work presents a numerical and experimental investigation on the heat transfer and turbulent flow of cooling steam in a rectangular duct with 90 .deg. ribs and studies the effect of cooling conditions on the heat transfer augmentation of steam. In the calculation, the variation range of Reynolds is from 10,000 to 190,000, the inlet temperature varies from 300 .deg. C to 500 .deg. C and the outlet pressure is from 0.5MPa to 6MPa. The aforementioned wide ranges of flow parameters cover the actual operating condition of coolant used in the gas turbine blades. The computations are carried with four turbulence models (the standard k-ε, the renormalized group (RNG) k-ε, the Launder-Reece-Rodi (LRR) and the Speziale-Sarkar-Gatski (SSG) turbulence models). The comparison of numerical and experimental results reveals that the SSG turbulence model is suitable for steam flow in the ribbed duct. Therefore, adopting the conjugate calculation technique, further study on the steam heat transfer and flow characteristics is performed with SSG turbulence model. The results show that the variation of cooling condition strongly impacts the forced convection heat transfer of steam in the ribbed duct. The cooling supply condition of a relative low temperature and medium pressure could bring a considerable advantage on steam thermal enhancement. In addition, comparing the heat transfer level between steam flow and air flow, the performance advantage of using steam is also influenced by the cooling supply condition. Changing Reynolds number has little effect on the performance superiority of steam cooling. Increasing pressure would strengthen the advantage, but increasing temperature gives an opposite result.

  17. The effect of cooling conditions on convective heat transfer and flow in a steam-cooled ribbed duct

    Energy Technology Data Exchange (ETDEWEB)

    Shui, Linqi; Gao, Jianmin; Shi, Xiaojun; Liu, Jiazeng; Xu, Liang [Xi' an Jiaotong University, Xi' an (China)

    2014-01-15

    This work presents a numerical and experimental investigation on the heat transfer and turbulent flow of cooling steam in a rectangular duct with 90 .deg. ribs and studies the effect of cooling conditions on the heat transfer augmentation of steam. In the calculation, the variation range of Reynolds is from 10,000 to 190,000, the inlet temperature varies from 300 .deg. C to 500 .deg. C and the outlet pressure is from 0.5MPa to 6MPa. The aforementioned wide ranges of flow parameters cover the actual operating condition of coolant used in the gas turbine blades. The computations are carried with four turbulence models (the standard k-ε, the renormalized group (RNG) k-ε, the Launder-Reece-Rodi (LRR) and the Speziale-Sarkar-Gatski (SSG) turbulence models). The comparison of numerical and experimental results reveals that the SSG turbulence model is suitable for steam flow in the ribbed duct. Therefore, adopting the conjugate calculation technique, further study on the steam heat transfer and flow characteristics is performed with SSG turbulence model. The results show that the variation of cooling condition strongly impacts the forced convection heat transfer of steam in the ribbed duct. The cooling supply condition of a relative low temperature and medium pressure could bring a considerable advantage on steam thermal enhancement. In addition, comparing the heat transfer level between steam flow and air flow, the performance advantage of using steam is also influenced by the cooling supply condition. Changing Reynolds number has little effect on the performance superiority of steam cooling. Increasing pressure would strengthen the advantage, but increasing temperature gives an opposite result.

  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. Steam explosion - physical foundations and relation to nuclear reactor safety

    International Nuclear Information System (INIS)

    Schumann, U.

    1982-08-01

    'Steam explosion' means the sudden evaporation of a fluid by heat exchange with a hotter material. Other terms are 'vapour explosion', 'thermal explosion', and 'energetic fuel-coolant interaction (FCI)'. In such an event a large fraction of the thermal energy initially stored in the hot material may possibly be converted into mechanical work. For pressurized water reactors one discusses (e.g. in risk analysis studies) a core melt-down accident during which molten fuel comes into contact with water. In the analysis of the consequences one has to investigate steam explosions. In this report an overview over the state of the knowledge is given. The overview is based on an extensive literature review. The objective of the report is to provide the basic knowledge which is required for understanding of the most important theories on the process of steam explosions. Following topics are treated: overview on steam explosion incidents, work potential, spontaneous nucleation, concept of detonation, results of some typical experiments, hydrodynamic fragmentation of drops, bubbles and jets, coarse mixtures, film-boiling, scenario of a core melt-down accident with possible steam-explosion in a pressurized water reactor. (orig.) [de

  20. Design codes for fast reactor steam generators

    International Nuclear Information System (INIS)

    Townley, C.H.A.

    1978-01-01

    The paper reviews the design methods and design criteria which are available for fast reactor structures, and discusses the materials data which are required to demonstrate the integrity of the plant components. (author)

  1. Superheated steam annealing of pressurized water reactor vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1993-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

  2. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  3. The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

  4. Steam generator tubing development for commercial fast breeder reactors

    International Nuclear Information System (INIS)

    Sessions, C.E.; Uber, C.F.

    1981-01-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs

  5. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  6. Emergency cooling system with hot-water jet pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Reinsch, A.O.W.

    1977-01-01

    The ECCS for a PWR or BWR uses hot-water jet pumps to remove the thermal energy generated in the reactor vessel and stored in the water. The hot water expands in the nozzle part (Laval nozzle) of the jet pump and sucks in coolant (borated water) coming from a storage tank containing subcooled water. This water is mixing with the hot water/steam mixture from the Laval nozzle. The steam is condensed. The kinetic energy of the water is converted into a pressure increase which is sufficient to feed the water into the reactor vessel. The emergency cooling may further be helped by a jet condenser also operating according to the principle of a jet pump and condensing the steam generated in the reactor vessel. (DG) [de

  7. Steam leak detection in advance reactors via acoustics method

    International Nuclear Information System (INIS)

    Singh, Raj Kumar; Rao, A. Rama

    2011-01-01

    Highlights: → Steam leak detection system is developed to detect any leak inside the reactor vault. → The technique uses leak noise frequency spectrum for leak detection. → Testing of system and method to locate the leak is also developed and discussed in present paper. - Abstract: Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.

  8. 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.

  9. Two-dimensional modeling of water spray cooling in superheated steam

    Directory of Open Access Journals (Sweden)

    Ebrahimian Vahid

    2008-01-01

    Full Text Available Spray cooling of the superheated steam occurs with the interaction of many complex physical processes, such as initial droplet formation, collision, coalescence, secondary break up, evaporation, turbulence generation, and modulation, as well as turbulent mixing, heat, mass and momentum transfer in a highly non-uniform two-phase environment. While it is extremely difficult to systematically study particular effects in this complex interaction in a well defined physical experiment, the interaction is well suited for numerical studies based on advanced detailed models of all the processes involved. This paper presents results of such a numerical experiment. Cooling of the superheated steam can be applied in order to decrease the temperature of superheated steam in power plants. By spraying the cooling water into the superheated steam, the temperature of the superheated steam can be controlled. In this work, water spray cooling was modeled to investigate the influences of the droplet size, injected velocity, the pressure and velocity of the superheated steam on the evaporation of the cooling water. The results show that by increasing the diameter of the droplets, the pressure and velocity of the superheated steam, the amount of evaporation of cooling water increases. .

  10. Radionuclide trap for liquid metal cooled reactors

    International Nuclear Information System (INIS)

    McGuire, J.C.; Brehm, W.F.

    1978-10-01

    At liquid metal cooled reactor operating temperatures, radioactive corrosion product transport and deposition in the primary system will be sufficiently high to limit access time for maintenance of system components. A radionuclide trap has been developed to aid in controlling radioactivity transport. This is a device which is located above the reactor core and which acts as a getter, physically immobilizing radioactive corrosion products, particularly 54 Mn. Nickel is the getter material used. It is most effective at temperatures above 450 0 C and effectiveness increases with increasing temperature. Prototype traps have been tested in sodium loops for 40,000 hours at reactor primary temperatures and sodium velocities. Several possible in-reactor trap sites were considered but a location within the top of each driver assembly was chosen as the most convenient and effective. In this position the trap is changed each time fuel is changed

  11. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  12. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

    1995-01-01

    The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

  13. Passive cooling systems in power reactors

    International Nuclear Information System (INIS)

    Aharon, J.; Harrari, R.; Weiss, Y.; Barnea, Y.; Katz, M.; Szanto, M.

    1996-01-01

    This paper reviews several R and D activities associated with the subject of passive cooling systems, conducted by the N.R.C.Negev thermohydraulic group. A short introduction considering different types of thermosyphons and their applications is followed by a detailed description of the experimental work, its results and conclusions. An ongoing research project is focused on the evaluation of the external dry air passive containment cooling system (PCCS) in the AP-600 (Westinghouse advanced pressurized water reactor). In this context some preliminary theoretical results and planned experimental research are for the fature described

  14. Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

    International Nuclear Information System (INIS)

    Foerster, S.; Hewing, G.

    1977-01-01

    On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

  15. Questions raised in developing fast reactor steam generator designs

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, P A; Hayden, O

    1975-07-01

    The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

  16. Questions raised in developing fast reactor steam generator designs

    International Nuclear Information System (INIS)

    Taylor, P.A.; Hayden, O.

    1975-01-01

    The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

  17. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  18. Startup of the FFTF sodium cooled reactor

    International Nuclear Information System (INIS)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed

  19. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  20. 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

  1. Emergency systems and protection equipment of modular steam generators for fast reactors

    International Nuclear Information System (INIS)

    Matal, O.

    The requirements are discussed for accident protection of modular steam generators for fast reactors. Accident protection is assessed for a modular through-flow steam generator and for a natural circulation modular steam generator. Benefits and constraints are shown and possible improvements are outlined for accident protection of liquid sodium fired modular steam generators. (Kr)

  2. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

    Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

  3. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

    Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

  4. 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

  5. Steam generators: critical components in nuclear steam supply systems

    Energy Technology Data Exchange (ETDEWEB)

    Stevens-Guille, P D

    1974-02-28

    Steam generators are critical components in power reactors. Even small internal leaks result in costly shutdowns for repair. Surveys show that leaks have affected one half of all water-cooled reactors in the world with steam generators. CANDU reactors have demonstrated the highest reliability. However, AECL is actively evolving new technology in design, manufacture, inspection and operation to maintain reliability. (auth)

  6. Thermal and flow design of helium-cooled reactors

    International Nuclear Information System (INIS)

    Melese, G.; Katz, R.

    1984-01-01

    This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors

  7. A resting bottom sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Costes, D.

    2012-01-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  8. HELB Analysis for ESBWR Reactor Building and Main Steam Tunnel

    Energy Technology Data Exchange (ETDEWEB)

    Noguera Oliva, O.

    2011-07-01

    The Reactor Building compartments and tbe Main Steam Tunnel are modeled using GOTHIC 7.2a. These models are based on Control Volumes (Rooms/Compartments/Regions), Flow Paths (junctions such as vent path or any opening) and Boundary Conditions (Mass and energy releases and outside conditions). Due to the different break locations, four models are built to analyze the short-term pressurization response. Are shown the cases analyzed, the results obtained and the models used for this purpose.

  9. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  10. Contrastive experimental study on heat transfer and friction characteristics in steam cooled and air cooled rectangular channels with rib turbulators

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Jianying; Li, Guojun; Gao, Tieyu [Xian Jiaotong University, Xian (China)

    2014-09-15

    The present experiment compares the heat transfer and friction characteristics in steam cooled and air cooled rectangular channels (simulating a gas turbine blade cooling passage) with two opposite rib-roughened walls. The Reynolds number (Re) whose length scale is the hydraulic diameter of the passage is set within the range of 10000-60000. The channel length is 1000 mm. The pitch-to-rib height ratio, the channel aspect ratio and the channel blockage ratio is 10, 0.5 and 0.047, respectively. It is found that the average Nu, the average friction coefficient, and the heat transfer performance of both steam and air in the ribbed channels show almost the same change trend with the increase of Re. Under the same test conditions, the average Nu of steam is 30.2% higher than that of air, the average friction coefficient is 18.4% higher, and the heat transfer performances of steam on the ribbed and the smooth walls are 8.4% and 7.3% higher than those of air, respectively. In addition, semi-empirical correlations for the two test channels are developed, which can predict the Nu under the given test condition. The correlations can be used in the design of the internal cooling passage of new generation steam cooled gas turbine blade/vane.

  11. Steam relief valve control system for a nuclear reactor

    International Nuclear Information System (INIS)

    Torres, J.M.

    1976-01-01

    Described is a turbine follow system and method for Pressurized Water Reactors utilizing load bypass and/or atmospheric dump valves to provide a substitute load upon load rejection by bypassing excess steam to a condenser and/or to the atmosphere. The system generates a variable pressure setpoint as a function of load and applies an error signal to modulate the load bypass valves. The same signal which operates the bypass valves actuates a control rod automatic withdrawal prevent to insure against reactor overpower

  12. Methanol steam-reforming in a catalytic fixed bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duesterwald, H G; Hoehlein, B; Kraut, H; Meusinger, J; Peters, R [Research Centre Juelich (KFA) (Germany). Inst. of Energy Process Engineering; Stimming, U [Technische Univ. Muenchen, Garching (Germany). Inst. fuer Festkoerperphysik und Techn. Phys.

    1997-12-01

    Designing an appropriate methanol steam reformer requires detailed knowledge about the processes within such a reactor. Thus, the axial temperature and concentration gradients and catalyst ageing were investigated. It was found that for a fresh catalyst load, the catalyst located in the reactor entrance was most active during the experiment. The activity of this part of the catalyst bed decreased after some time of operation due to ageing. With further operation, the most active zone moved through the catalyst bed. From the results concerning hydrogen production and catalyst degradation, the necessary amount of catalyst for a mobile PEMFC-system can be estimated. (orig.)

  13. Steam reforming of heptane in a fluidized bed membrane reactor

    Science.gov (United States)

    Rakib, Mohammad A.; Grace, John R.; Lim, C. Jim; Elnashaie, Said S. E. H.

    n-Heptane served as a model compound to study steam reforming of naphtha as an alternative feedstock to natural gas for production of pure hydrogen in a fluidized bed membrane reactor. Selective removal of hydrogen using Pd 77Ag 23 membrane panels shifted the equilibrium-limited reactions to greater conversion of the hydrocarbons and lower yields of methane, an intermediate product. Experiments were conducted with no membranes, with one membrane panel, and with six panels along the height of the reactor to understand the performance improvement due to hydrogen removal in a reactor where catalyst particles were fluidized. Results indicate that a fluidized bed membrane reactor (FBMR) can provide a compact reformer for pure hydrogen production from a liquid hydrocarbon feedstock at moderate temperatures (475-550 °C). Under the experimental conditions investigated, the maximum achieved yield of pure hydrogen was 14.7 moles of pure hydrogen per mole of heptane fed.

  14. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    Science.gov (United States)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  15. Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

  16. Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

    International Nuclear Information System (INIS)

    Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum

    2005-01-01

    KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types

  17. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  18. Acoustic noises of the BOR-60 reactor steam generators when simulating leaks with argon and steam

    International Nuclear Information System (INIS)

    Sokolov, V.M.; Golushko, V.V.; Afanas'ev, V.A.; Grebenkin, Yu.P.; Muralev, A.B.

    1985-01-01

    Background acoustic noises of stea generators in different operational regimes and noises of argon and steam small leads (about 0.1 g/s) are studied to determine the possibility of designing the acoustic system for leak detection in sodium-water steamgenerators. Investigations are carried out at the 30 MW micromodule steam generator being in operation at the BOR-60 reactor as well as at the 20 MW tank type steam generator. Immersed ransduceres made of lithium niobate 6 mm in-diameter and waveguide transducers made of a stainless steel in the form of rods 10 mm in-diameter and 500 mm long are used as acoustic monitors. It is shown that the leak noise is more wide-band than the background noise of the steam generator and both high and low frequencies appear in the spectrum. The use of monitors of different types results in similar conslusions inrelation to the character of background noises and leak signals (spectral density, signal to-noise ratio) in the ase of similar bandroidths of the transduceres. A conclusion is made that the change of operational regimes leads to changes of background noise level, which can be close to the reaction of

  19. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  20. 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

  1. 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.

  2. MEANS FOR SHIELDING AND COOLING REACTORS

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

    1959-02-10

    Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

  3. Bacterial pathogens in a reactor cooling reservoir

    International Nuclear Information System (INIS)

    Kasweck, K.L.; Fliermans, C.B.

    1978-01-01

    The results of the sampling in both Par Pond and Clark Hill Reservoir are given. The frequency of isolation is a qualitative parameter which indicates how often the specified bacterium was isolated from each habitat. Initial scoping experiments demonstrated that a wider variety of pathogenic bacteria occur in Par Pond than in Clark Hill Reservoir. Such findings are interesting because Par Pond does not receive any human wastes directly, yet bacteria generally associated with human wastes are more frequently isolated from Par Pond. Previous studies have demonstrated that certain non-spore-forming enteric bacteria do not survive the intense heat associated with the cooling water when the reactor is operating. However, even when the reactor is not operating, cooling water, consisting of 10% makeup water from Savannah River, continues to flow into Par Pond. This flow provides a source of bacteria which inoculate Par Pond. Once the reactor is again operating, these same bacteria appear to be able to survive and grow within the Par Pond system. Thus, Par Pond and the associated lakes and canals of the Par Pond system provide a pool of pathogens that normally would not survive in natural waters

  4. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

    Mitake, Susumu; Suzuki, Katsuo; Miyamoto, Yoshiaki; Tamura, Kazuo; Ezaki, Masahiro.

    1983-03-01

    Performances and design of the panel cooling system which has been proposed to be equipped as an emergency reactor cooling system for the experimental multi purpose very high temperature gas-cooled reactor are explained. Effects of natural circulation flow which would develop in the core and temperature transients of the panel in starting have been precisely investigated. Conditions and procedures for settling accidents with the proposed panel cooling system have been also studied. Based on these studies, it has been shown that the panel cooling system is effective and useful for the emergency reactor cooling of the experimental VHTR. (author)

  5. 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.)

  6. Hybrid dry cooling. Enhanced energy efficiency in steam turbines. Hybride Trockenkuehlung. Bessere Energienutzung bei Dampfturbinen

    Energy Technology Data Exchange (ETDEWEB)

    Stucki, R. (Colenco Power Consulting AG, Baden (Switzerland)); Mueller, W. (Jaeggi AG, Bern (Switzerland)); Haltiner, E.W.

    1992-09-11

    Condensation turbo groups facilitate the utilization of thermal energy from vapor production plants, for example from garbage incineration. Through an optimum design of the cooling system, the profitable electric energy can be increased through the maximum spreading of the condensation temperature and the live steam temperature. The hybrid dry cooling guarantees unsurpassable operating conditions. (orig.).

  7. Screening reactor steam/water piping systems for water hammer

    International Nuclear Information System (INIS)

    Griffith, P.

    1997-09-01

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

  8. 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

  9. 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.

  10. Selection of steam generator materials for sodium cooled fast breeders

    International Nuclear Information System (INIS)

    Berge, P.

    1977-01-01

    The sodium water heat exchangers are now considered as the stumbling block in the development of liquid metal cooled fast breeders, due to the risk of sodium-water reactions. The selection of the materials for these tube-bundles has been very broad, for the different existing, or in-project, reactors in the world: low alloy 2 1/4 Cr - 1 Mo steels (unstabilized or stabilized); 9 Cr - 1 Mo ferritic steel; 18 Cr - 10 Ni austenitic stainless steels; alloy 800. On can also add other ferritic steels, as 9 Cr - 2 Mo stabilized, which are studied for this application. In the framework of the E.D.F.-C.E.A. working group a major effort was undertaken to study the characteristics of these various materials with respect to the main criteria governing construction of the tube bundles and their performance in service: mechanical characteristics at high temperature; fabrication and welding; behavior with respect to mass transfer in sodium; carburization and decarburization; corrosion resistance. The main lines and results of this program are described [fr

  11. Hydrogen detector for sodium cooled reactors

    International Nuclear Information System (INIS)

    Roy, P.; Rodgers, D.N.

    1975-01-01

    An improved hydrogen detector for use in sodium cooled reactors is described. The improved detector basically comprises a diffusion tube of either pure nickel or stainless steel having a coating on the vacuum side (inside) of a thin layer of refractory metal, e.g., tungsten or molybdenum. The refractory metal functions as a diffusion barrier in the path of hydrogen diffusing from the sodium on the outside of the detector into the vacuum on the inside, thus by adjusting the thickness of the coating, it is possible to control the rate of permeation of hydrogen through the tube, thereby providing a more stable detector. (U.S.)

  12. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    Aspden, G.J.; Allbeson, K.F.

    1984-01-01

    In a liquid metal cooled nuclear reactor with a nuclear fuel assembly in a coolant-containing primary vessel housed within a concrete containment vault, there is thermal insulation to protect the concrete, the insulation being disposed between vessel and concrete and being hung from metal structure secured to and projecting from the concrete, the insulation consisting of a plurality of adjoining units each unit incorporating a pack of thermal insulating material and defining a contained void co-extensive with said pack and situated between pack and concrete, the void of each unit being connected to the voids of adjoining units so as to form continuous ducting for a fluid coolant. (author)

  13. Emergency cooling system for the PHENIX reactor

    International Nuclear Information System (INIS)

    Megy, J.M.; Giudicelli, A.G.; Robert, E.A.; Crette, J.P.

    Among various engineered safeguards of the reactor plant, the authors describe the protective system designed to remove the decay heat in emergency, in case of complete loss of all normal decay heat removal systems. First the normal decay heat rejection systems are presented. Incidents leading to the loss of these normal means are then analyzed. The protective system and its constructive characteristics designed for emergency cooling and based on two independent and highly reliable circuits entirely installed outside the primary containment vessel are described

  14. Saturated steam turbines for power reactors of WWER-type

    International Nuclear Information System (INIS)

    Czwiertnia, K.

    1978-01-01

    The publication deals with design problems of large turbines for saturated steam and with problem of output limitations of single shaft normal speed units. The possibility of unification of conventional and nuclear turbines, which creates the economic basis for production of both types of turbines by one manufacturer based on standarized elements and assemblies is underlined. As separate problems the distribution of nuclear district heating power systems are considered. The choice of heat diagram for district heating saturated steam turbines, the advantages of different diagrams and evaluaton for further development are presented. On this basis a program of unified turbines both condensing and district heating type suitable for Soviet reactors of WWER-440 and WWER-1000 type for planned development of nuclear power in Poland is proposed. (author)

  15. 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.

  16. 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.)

  17. Design features of Advanced Power Reactor (APR) 1400 steam generator

    International Nuclear Information System (INIS)

    Park, Tae-Jung; Park, Jun-Soo; Kim, Moo-Yong

    2004-01-01

    Advanced Power Reactor 1400 (APR 1400) which is to achieve the improvement of the safety and economical efficiency has been developed by Korea Hydro and Nuclear Power Co., Ltd. (KHNP) with the support from industries and research institutes. The steam generator for APR 1400 is an evolutionary type from System 80 + , which is the recirculating U-tube heat exchanger with integral economizer. Compared to the System 80 + steam generator, it is focused on the improved design features, operating and design conditions of APR 1400 steam generator. Especially, from the operation experience of Korean Standard Nuclear Power Plant (KSNP) steam generator, the lessons-learned measures are incorporated to prevent the tube wear caused by flow-induced vibration (FIV). The concepts for the preventive design features against FIV are categorized to two fields; flow distribution and dynamic response characteristics. From the standpoint of flow distribution characteristics, the egg-crate flow distribution plate (EFDP) is installed to prevent the local excessive flow loaded on the most susceptible tube to wear. The parametric study is performed to select the optimum design with the efficient mitigation of local excessive flow. ATHOS3 Mod-01 is used and partly modified to analyze the flow field of the APR 1400 steam generator. In addition, the upper tube bundle support is designed to eliminate the presence of tube with a low natural frequency. Based on the improved upper tube bundle support, the modal analysis is performed and compared with that of System 80 + . Using the results of flow distribution and modal analysis, the two mechanisms of flow-induced vibration are investigated; fluid-elastic instability (FEI) and random turbulence excitation (RTE). (authors)

  18. 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

  19. Study on thermodynamic cycle of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Qu Xinhe; Yang Xiaoyong; Wang Jie

    2017-01-01

    The development trend of the (very) High temperature gas-cooled reactor is to gradually increase the reactor outlet temperature. The different power conversion units are required at the different reactor outlet temperature. In this paper, for the helium turbine direct cycle and the combined cycle of the power conversion unit of the High temperature gas-cooled reactor, the mathematic models are established, and three cycle plans are designed. The helium turbine direct cycle is a Brayton cycle with recuperator, precooler and intercooler. In the combined cycle plan 1, the topping cycle is a simple Brayton cycle without recuperator, precooler and intercooler, and the bottoming cycle is based on the steam parameters (540deg, 6 MPa) recommended by Siemens. In the combined cycle plan 2, the topping cycle also is a simple Brayton cycle, and the bottoming cycle which is a Rankine cycle with reheating cycle is based on the steam parameters of conventional subcritical thermal power generation (540degC, 18 MPa). The optimization results showed that the cycle efficiency of the combined cycle plan 2 is the highest, the second is the helium turbine direct cycle, and the combined cycle plan 2 is the lowest. When the reactor outlet temperature is 900degC and the pressure ratio is 2.02, the cycle efficiency of the combined cycle plan 2 can reach 49.7%. The helium turbine direct cycle has a reactor inlet temperature above 500degC due to the regenerating cycle, so it requires a cooling circuit for the internal wall of the reactor pressure vessel. When the reactor outlet temperature increases, the increase of the pressure ratio required by the helium turbine direct cycle increases may bring some difficulties to the design and manufacture of the magnetic bearings. For the combined cycle, the reactor inlet temperature can be controlled below than 370degC, so the reactor pressure vessel can use SA533 steel without cooling the internal wall of the reactor pressure vessel. The pressure

  20. Renewal of reactor cooling system of JMTR. Reactor building site

    International Nuclear Information System (INIS)

    Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

    2012-03-01

    The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

  1. Steam drum level control studies of a natural circulation multi loop reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

    2013-12-15

    The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

  2. Steam generator performance improvements for integral small modular reactors

    Directory of Open Access Journals (Sweden)

    Muhammad Ilyas

    2017-12-01

    Results and Conclusions: The results are compared with helical-coiled SGs being used in IRIS (International Reactor Innovative and Secure. The results show that the tube length reduces up to 11.56% keeping thermal and hydraulic conditions fixed. In the case of fixed size, the steam outlet temperature increases from 590.1 K to 597.0 K and the capability of power transfer from primary to secondary also increases. However, these advantages are associated with some extra pressure drop, which has to be compensated.

  3. Twin header bore welded steam generator for pressurized water reactors

    International Nuclear Information System (INIS)

    Davies, R.J.; Hirst, B.

    1979-01-01

    A description is given of a pressurized water reactor (PWR) steam generator concept, several examples of which have been in service for up to fourteen years. Details are given of the highly successful service record of this equipment and the features which have been incorporated to minimize corrosion and deposition pockets. The design employs a vertical U tube bundle carried off two horizontal headers to which the tubes are welded by the Foster Wheeler Power Products (FWPP) bore welding process. The factors to be considered in uprating the design to meet the current operating conditions for a 1000 MW unit are discussed. (author)

  4. Calandria cooling structure in pressure tube reactor

    International Nuclear Information System (INIS)

    Hyugaji, Takenori; Sasada, Yasuhiro.

    1976-01-01

    Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

  5. 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

  6. European supercritical water cooled reactor (HPLWR Phase 2 project)

    International Nuclear Information System (INIS)

    Schulenberg, Thomas; Starflinger, Joerg; Marsault, Philippe; Bittermann, Dietmar; Maraczy, Czaba; Laurien, Eckart; Lycklama, Jan Aiso; Anglart, Henryk; Andreani, Michele; Ruzickova, Mariana; Heikinheimo, Liisa

    2010-01-01

    The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 deg C maximum core outlet temperature. It is designed and analyzed by a European consortium of 13 partners from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small, housed fuel assemblies with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The innovative core design with upward and downward flow through its assemblies has been studied with neutronic, thermal-hydraulic and stress analyses and has been reviewed carefully in a mid-term assessment. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. An overview of results achieved up to now, given in this paper, is illustrating the latest scientific and technological advances. (author)

  7. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

    Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy; Amit Jain Han Sang Kim; Vishisht Gupta; Jonathan Pitt

    2006-01-01

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

  8. Specific features of phase distribution in a draught part of the tank type boiling water cooled reactor

    International Nuclear Information System (INIS)

    Fedulin, V.N.; Bartolomej, G.G.; Solodkij, V.A.; Shmelev, V.E.

    1984-01-01

    The results of experimental investigation of the two-phase flow structure in a draught part of the VK-50 boiling water cooled reactor are presented. A qualitative physical model of steam-water mixture flow in the large diameter draught part is suggested. It is shown that for hydrodynamically unstable two-phase flows a considerable nonuniformity in steam content distribution over the draught part volume which determines the possibility of the recirculating coolant flow formation in the peripheral zone is observed. At the draught part inlet the radial distribution of steam content is determined by the complex effects of power distribution and coolant flow rate change over the core radius. The flow structure in the lower section of the draught part adjoining to the core is determined to a considerable degree by a coolant jet outflow from fuel assembly (FA) nozzels Jet height depends on the velocity of outgoing two-phase flow, working pressure and hydrodynamics of the draught part. The jet height does not exceed 0.4 m for the K-50 reactor. Due to the increased steam outflow from the central FAs and the existence of radial pressure gradient the water-steam mixture is turned from the draught part periphery to its central part, where accelerated water steam flow with an increased steam content is formed. When a certain height is achieved a graduel expansion of the water-steam flow begins leading to equalizing the steam content over the draught part cross section

  9. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  10. Integration of steam injection and inlet air cooling for a gas turbine generation system

    International Nuclear Information System (INIS)

    Wang, F.J.; Chiou, J.S.

    2004-01-01

    The temperature of exhaust gases from simple cycle gas turbine generation sets (GENSETs) is usually very high (around 500 deg. C), and a heat recovery steam generator (HRSG) is often used to recover the energy from the exhaust gases and generate steam. The generated steams can be either used for many useful processes (heating, drying, separation etc.) or used back in the power generation system for enhancing power generation capacity and efficiency. Two well-proven techniques, namely steam injection gas turbine (STIG) and inlet air cooling (IAC) are very effective features that can use the generated steam to improve the power generation capacity and efficiency. Since the energy level of the generated steam needed for steam injection is different from that needed by an absorption chiller to cool the inlet air, a proper arrangement is required to implement both the STIG and the IAC features into the simple cycle GENSET. In this study, a computer code was developed to simulate a Tai power's Frame 7B simple cycle GENSET. Under the condition of local summer weather, the benefits obtained from the system implementing both STIG and IAC features are more than a 70% boost in power and 20.4% improvement in heat rate

  11. The steam explosion potential for an unseated SRS reactor septifoil

    International Nuclear Information System (INIS)

    Allison, D.K.; Hyder, M.L.; Yau, W.W.F.; Smith, D.C.

    1992-01-01

    Control rods in the Savannah River Site's K Reactor are contained within housings composed of seven channels (''septifoils''). Each septifoil is suspended from the top of the reactor and is normally seated on an upflow pin that channels coolant to the septifoil. Forced flow to the septifoil would be eliminated in the unlikely event of a septifoil unseated upon installation, i.e., if the septifoil is not aligned with its upflow pin. If this event were not detected, control rod melting and the interaction of molten metal with water might occur. This paper describes a methodology used to address the issue of steam explosions that might arise by this mechanism. The probability of occurrence of a damaging steam explosion given an unseated septifoil was found to be extremely low. The primary reasons are: (1) the high probability that melting will not occur, (2) the possibility of material holdup by contact with the outer septifoil housing, (3) the relative shallowness of the pool 'Of water into which molten material might fall, (4) the probable absence of a trigger, and (5) the relatively large energy release required to damage a nearby fuel assembly. The methodology is based upon the specification of conditions prevailing within the septifoil at the time molten material is expected to contact water, and upon information derived from the available experimental data base, supplemented by recent prototypic experiments

  12. A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

    International Nuclear Information System (INIS)

    Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

    2015-01-01

    A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

  13. Simplified numerical simulation of hot channel in sodium cooled reactor

    International Nuclear Information System (INIS)

    Fonseca, F. de A.S. da; Silva Filho, E.

    1988-12-01

    The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author) [pt

  14. A passive emergency heat sink for water cooled reactors with particular application to CANDU reg-sign reactors

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU reg-sign moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners

  15. 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

  16. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  17. Technological readiness of evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Juhn, P.E.

    1999-01-01

    Nuclear energy has evolved to a mature industry that supplies over 16% of the world's electricity, and it represents an important option for meeting the global energy demands of the coming century in an environmentally acceptable manner. New, evolutionary water cooled reactor designs that build on successful performance of predecessors have been developed; these designs have generally been guided by wishes to reduce cost, to improve availability and reliability, and to meet increasingly stringent safety objectives. These three aspects are important factors in what has been called technological readiness for an expanded deployment of nuclear power; a major increase in utilization of nuclear power will only occur if it is economically competitive, and meets safety expectations. To this end, the industry will also have to maintain or improve the public perception of nuclear power as a benign, economical and reliable energy source. (author)

  18. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Duncombe, E.; Thatcher, G.

    1979-01-01

    The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

  19. 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

  20. 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

  1. Protected air-cooled condenser for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Louison, R.; Boardman, C.E.

    1981-01-01

    The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events

  2. Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

    International Nuclear Information System (INIS)

    Tomabechi, K.

    1975-01-01

    The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

  3. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  4. Chooz A: a model for the dismantling of water-cooled reactors

    International Nuclear Information System (INIS)

    Anon.

    2017-01-01

    The specificity of Chooz-A, the first French pressurized water reactor (PWR), is that the reactor and its major components (pumps, exchangers and cooling circuits) are installed in 2 caves dug out in a hill slope. Chooz-A was operating from 1967 to 1991, in 1993 the fuel was removed and in 2007 EDF received the authorization to dismantle the reactor. In 2012, EDF completed the dismantling of the cave containing the elements of the cooling circuit, a cornerstone was the removing of the four 14 m high steam generators. The dismantling of the pressure vessel began in march 2017, it is the same tools and the same processes that were used for the dismantling of the pressure vessel of the Zorita plant (Spain) in 2016. The end of the Chooz-A dismantling is expected in 2022. The feedback experience will help to standardize practices for the French fleet of PWRs. (A.C.)

  5. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    Improvements in the design of the thermally insulating material used to shield the concrete containment walls in liquid metal cooled nuclear reactors are described in detail. The insulating material is composed of two layers and is placed between the primary vessel (usually steel) and the steel lined concrete containment vault. The outer layer, which clads the inner wall surface of the vault, is generally impervious to liquid metal coolant whilst the inner layer is pervious to the coolant. In normal operation, both layers protect the concrete from heat radiated from the reactor. In the event of a breach of the containment vessel, the resulting leakage of liquid metal coolant permeates the inner layer of insulating material, provides a means of heat transfer by conduction and hence reduces the overall insulating properties of the two layers. The outer layer continues to protect the wall surface of the vault from substantial direct contact with the liquid metal. Thus the two apparently conflicting requirements of good thermal insulation during normal operation and of heat transfer during loss of coolant accidents are satisfied by this novel design. Suggestions are given for possible materials for use as the insulating layers. (U.K.)

  6. Technical evaluation of corium cooling at the reactor cavity

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Chan, Eun Sun; Lee, Jae Hun; Lee, Jong In

    1998-01-01

    To terminate the progression of the severe accident and mitigate the accident consequences, corium cooling has been suggested as one of most important design features considered in the severe accident mitigation. Till now, some kinds of cooling methodologies have been identified and, specially, the corium cooling at the reactor cavity has been considered as one of the most promising cooling methodologies. Moreover, several design requirements related to the corium cooling at the reactor cavity have been also suggested and applied to the design of the next generation reactor. In this study, technical descriptions are briefly described for the important issues related to the corium cooling at the reactor cavity, i.e. cavity area, cavity flooding system, etc., and simple evaluations for those items have been performed considering present technical levels including the experiment and analytical works

  7. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1976-01-01

    Reference is made to liquid metal cooled fast breeder reactors of the 'pool' kind. In this type of reactor the irradiated fuel is lowered into a transfer rotor for removal to storage facilities, this rotor normally having provision for the temporary storage of 20 irradiated fuel assemblies, each within a stainless steel bucket. For insertion or withdrawal of a fuel assembly the rotor is rotated to bring the fuel assembly to a loading or discharging station. The irradiated fuel assembly is withdrawn from the rotor within its bucket and the total weight is approximately 1000 kg, which is lifted about 27 m. In the event of malfunction the combination falls back into the rotor with considerable force. In order to prevent damage to the rotor fracture pins are provided, and to prevent damage to the reactor vessel and other parts of the reactor structure deformable energy absorbing devices are provided. After a malfunction the fractured pins and the energy absorbing devices must be replaced by remote control means operated from outside the reactor vault - a complex operation. The object of the arrangement described is to provide improved energy absorbing means for fuel assemblies falling into a fuel transfer rotor. The fuel assemblies are supported in the rotor by elastic means during transfer to storage and a hydraulic dash pot is provided in at least one position below the rotor for absorbing the energy of a falling fuel assembly. It is preferable to provide dash pots immediately below a receiving station for irradiated fuel assemblies and immediately below a discharge station. Each bucket is carried in a container that is elastically supported in the transfer rotor on a helical coil compression spring, so that, in the event of a malfunction the container and bucket are returned to their normal operating position after the force of the falling load has been absorbed by the dash pot. The transfer rotor may also be provided with recoil springs to absorb the recoil energy

  8. Natural draft dry-type cooling tower for steam power plants

    International Nuclear Information System (INIS)

    Nasser, G.

    1976-01-01

    The task to build natural-draught dry cooling towers for large steam power plants as simple, compact, and economical as possible may be achieved by a combination of known features with the aid of the present application: the condenser elements built as piles of corrugated plates are arranged in the form of a truncated pyramid widened towards the top. For the cooling-air flow inlet openings for hot gas supplied from the lower part of the dome are provided. (UWI) [de

  9. A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

    International Nuclear Information System (INIS)

    Yoshihisa Shindo; Kazuo Haga

    2005-01-01

    Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

  10. Thermosyphon Phenomenon as an alternate heat sink of Shutdown Cooling System for the CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jonghyun [GNEST, Seoul (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

    2006-07-01

    During the outage(overhaul) of the CANDU plant, there is a period when the coolant is partially drained to the reactor header level and the coolant is cooled and depressurized by Shutdown Cooling System(SDCS) other than PHTS pump. In the postulated accident of the loss of SDCS-the PHTS pump failure, the primary coolant system should be cooled by the alternate heat sink using the thermosyphon pheonomenon(TS) through the steam generator(SG) This study was aimed at verification and analyzing the core cooling ability of the TS. And the sensitivity analysis was done for the number of SGs used in the TS. As an analysis tool, RELAP5/CANDU was used.

  11. Passive containment cooling system performance in the simplified boiling water reactor

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Gamble, R.E.; Yadigaroglu, G.

    1997-01-01

    The Simplified Boiling Water Reactor (SBWR) incorporates a passive system for decay heat removal from the containment in the event of a postulated Loss-of-Coolant Accident (LOCA). Decay heat is removed by condensation of the steam discharged from the reactor pressure vessel (RPV) in three condensers which comprise the Passive Containment Cooling System (PCCS). These condensers are designed to carry the heat load while transporting a mixture of steam and noncondensible gas (primarily nitrogen) from the drywell to the suppression chamber. This paper describes the expected LOCA response of the SBWR with respect to the PCCS performance, based on analysis and test results. The results confirm that the PCCS has excess capacity for decay heat removal and that overall system performance is very robust. 12 refs., 8 figs

  12. 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)

  13. Development Project of Supercritical-water Cooled Power Reactor

    International Nuclear Information System (INIS)

    Kataoka, K.; Shiga, S.; Moriya, K.; Oka, Y.; Yoshida, S.; Takahashi, H.

    2002-01-01

    A Supercritical-water Cooled Power Reactor (SCPR) development project (Feb. 2001- Mar. 2005) is being performed by a joint team consisting of Japanese universities and nuclear venders with a national fund. The main objective of this project is to provide technical information essential to demonstration of SCPR technologies through concentrating three sub-themes: 'plant conceptual design', 'thermohydraulics', and 'material and water chemistry'. The target of the 'plant conceptual design sub-theme' is simplify the whole plant systems compared with the conventional LWRs while achieving high thermal efficiency of more than 40 % without sacrificing the level of safety. Under the 'thermohydraulics sub-theme', heat transfer characteristics of supercritical-water as a coolant of the SCPR are examined experimentally and analytically focusing on 'heat transfer deterioration'. The experiments are being performed using fron-22 for water at a fossil boiler test facility. The experimental results are being incorporated in LWR analytical tools together with an extended steam/R22 table. Under the 'material and water chemistry sub-theme', material candidates for fuel claddings and internals of the SCPR are being screened mainly through mechanical tests, corrosion tests, and simulated irradiation tests under the SCPR condition considering water chemistry. In particular, stress corrosion cracking sensitivity is being investigated as well as uniform corrosion and swelling characteristics. Influences of water chemistry on the corrosion product characteristics are also being examined to find preferable water condition as well as to develop rational water chemistry controlling methods. (authors)

  14. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  15. Control of Canadian once-through direct cycle supercritical water-cooled reactors

    International Nuclear Information System (INIS)

    Sun, Peiwei; Wang, Baosheng; Zhang, Jianmin; Su, Guanghui

    2015-01-01

    Highlights: • Dynamic characteristics of Canadian SCWR are analyzed. • Hybrid feedforward and feedback control is adopted to deal with cross-coupling. • Gain scheduling control with smooth weight is applied to deal with nonlinearity. • It demonstrates through simulation that the control requirements are satisfied. - Abstract: Canadian supercritical water-cooled reactor (SCWR) can be modelled as a Multiple-input Multiple-output (MIMO) system. It has a high power-to-flow ratio, strong cross-coupling and high degree of nonlinearity in its dynamic characteristics. Among the outputs, the steam temperature is strongly affected by the reactor power and the most challenging to control. It is difficult to adopt a traditional control system design methodology to obtain a control system with satisfactory performance. In this paper, feedforward control is applied to reduce the effect on steam temperature from the reactor power. Single-input Single-output (SISO) feedback controllers are synthesized in the frequency domain. Using the feedforward controller, the steam temperature variation due to disturbances at the reactor power has been significantly suppressed. The control system can effectively maintain the overall system stability and regulate the plant around a specified operating condition. To deal with the nonlinearities, gain scheduling control strategy is adopted. Different sets of controllers combined by smooth weight functions are used for the plant at different load conditions. The proposed control strategies have been evaluated under various operating scenarios. Simulation results show that satisfactory performance can successfully achieved by the designed control system

  16. Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Scherer, Winfried

    1997-01-01

    This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

  17. Reactor core of light water-cooled reactor

    International Nuclear Information System (INIS)

    Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

    1996-01-01

    In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

  18. Design considerations for economically competitive sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

    2009-01-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

  19. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  20. Nuclear reactor lid cooling which can work by natural circulation

    International Nuclear Information System (INIS)

    Wagner, J.

    1985-01-01

    The well-known air cooling of the lid of liquid metal cooled nuclear reactors is improved by the start of natural convection flow ensuring removal of heat in a sufficiently short time, if the blower fails. Go and return branches of the individual cooling circuits are arranged at different heights for this purpose. The circulation is supported by opening valves, which provide a direct path into the reactor building for the cooling air. The draught can be increased by setting up special chimneys. The start of circulation is aided by the temporary opening of another valve. (orig.) [de

  1. Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

    International Nuclear Information System (INIS)

    Permana, Sidik

    2009-01-01

    Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

  2. Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Scherer, W.

    1996-01-01

    This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H 2 O concentration increase in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduce the ingress velocity of the H 2 O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H 2 O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [de

  3. Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

    Science.gov (United States)

    Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

  4. Uranium utilization of light water cooled reactors and fast breeders

    International Nuclear Information System (INIS)

    Stojadinovic, Timm

    1991-08-01

    The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

  5. Passive cooling of a fixed bed nuclear reactor

    International Nuclear Information System (INIS)

    Petry, V.J.; Bortoli, A.L. de; Sefidwash, F.

    2005-01-01

    Small nuclear reactors without the need for on-site refuelling have greater simplicity, better compliance with passive safety systems, and are more adequate for countries with small electric grids and limited investment capabilities. Here the passive cooling characteristic of the fixed bed nuclear reactor (FBNR), that is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project, is studied. A mathematical model is developed to calculate the temperature distribution in the fuel chamber of the reactor. The results demonstrate the passive cooling of this nuclear reactor concept. (authors)

  6. Modeling and Simulation of the Multi-module High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Liu Dan; Sun Jun; Sui Zhe; Xu Xiaolin; Ma Yuanle; Sun Yuliang

    2014-01-01

    The modular high temperature gas-cooled reactor (MHTGR) is characterized with the inherent safety. To enhance its economic benefit, the capital cost of MHTGR can be decreased by combining more reactor modules into one unit and realize the batch constructions in the concept of modularization. In the research and design of the multi-module reactors, one difficulty is to clarify the coupling effects of different modules in operating the reactors due to the shared feed water and main steam systems in the secondary loop. In the advantages of real-time simulation and coupling calculations of different modules and sub-systems, the operation of multi-module reactors can be studied and analyzed to understand the range and extent of the coupling effects. In the current paper; the engineering simulator for the multi-module reactors was realized and able to run in high performance computers, based on the research experience of the HTR-PM engineering simulator. The models were detailed introduced including the primary and secondary loops. The steady state of full power operation was demonstrated to show the good performance of six-module reactors. Typical dynamic processes, such as adjusting feed water flow rates and shutting down one reactor; were also tested to study the coupling effects in multi-module reactors. (author)

  7. LOFA analyses for the water and helium cooled SEAFP reactors

    International Nuclear Information System (INIS)

    Sponton, L.; Sjoeberg, A.; Nordlinder, S.

    2001-01-01

    This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

  8. 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...

  9. Process for superheating the steam generated by a light water nuclear reactor

    International Nuclear Information System (INIS)

    Vakil, H.B.; Brown, D.H.

    1976-01-01

    A process is submitted for superheating the pressurised steam generated in a light water nuclear reactor in which the steam is brought to 340 0 C at least. This superheated steam is used to operate a turbo-generator unit. The characteristic of the process is that an exothermal chemical reaction is used to generate the heat utilised during the superheating stage. The chemical reaction is a mechanisation, oxidation-reduction or hydrogenation reaction [fr

  10. 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.)

  11. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Durston, J.G.

    1976-01-01

    It is stated that in a liquid metal cooled fast breeder reactor wherein the core, intermediate heat exchangers and liquid metal pumps are immersed in a pool of coolant such as Na, the intermediate heat exchangers are suspended from the roof, and ducting is provided in the form of a core tank or shroud interconnected with 'pods' housing the intermediate exchangers for directing coolant from the core over the heat exchanger tubes and thence back to the main pool of liquid metal. Seals are provided between the intermediate heat exchanger shells and the walls of their 'pods' to prevent liquid metal flow by-passing the heat exchanger tube bundles. As the heat exchangers must be withdrawable for servicing, and because linear differential thermal expansion of the heat exchanger and its 'pod' must be accommodated the seals hitherto have been of the sliding kind, generally known as 'piston ring type seals'. These present several disadvantages; for example sealing is not absolute, and the metal to metal seal gives rise to wear and fretting by rubbing and vibration. This could lead to seizure or jamming by the deposition of impurities in the coolant. Another difficulty arises in the need to accommodate lateral thermal expansion of the ducting, including the core tank and 'pods'. Hitherto some expansion has been allowed for by the use of expansible bellow pairs in the interconnections, or alternatively by allowing local deformations of the core tank 'pods'. Such bellows must be very flexible and hence constitute a weak section of the ducting, and local deformations give rise to high stress levels that could lead to premature failure. The arrangement described seeks to overcome these difficulties by use of a gas pocket trapping means to effect a seal against vertical liquid flow between the heat exchanger shell and the wall of the heat exchanger housing. Full details of the arrangement are described. (U.K.)

  12. Reactor-core isolation cooling system with dedicated generator

    International Nuclear Information System (INIS)

    Nazareno, E.V.; Dillmann, C.W.

    1992-01-01

    This patent describes a nuclear reactor complex. It comprises a dual-phase nuclear reactor; a main turbine for converting phase-conversion energy stored by vapor into mechanical energy for driving a generator; a main generator for converting the mechanical energy into electricity; a fluid reservoir external to the reactor; a reactor core isolation cooling system with several components at least some of which require electrical power. It also comprises an auxiliary pump for pumping fluid from the reservoir into the reactor pressure vessel; an auxiliary turbine for driving the pump; control means for regulating the rotation rate of the auxiliary turbine; cooling means for cooling the control means; and an auxiliary generator coupled to the auxiliary turbine for providing at least a portion of the electrical power required by the components during a blackout condition

  13. Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

    International Nuclear Information System (INIS)

    Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

    2008-01-01

    In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

  14. Studying the processes of sodium-water interaction in the BOR-60 reactor micromodule steam generator

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Antipin, G.K.; Borisov, V.V.

    1981-01-01

    Main results of experimental studies of emergency regimes of micromodule steam generator (MSG) at small and big leaks of water into sodium, realized using the 30 MW MSG, operating in the BOR-o0 reactor, are considered. The aims of the study are as follows: the modelling of macroleak in ''Nadja'' steam generator for the BN-350 reactor; testing the conceptions of alarm signalling and MSG protection; testing under real conditions of new perspective systems of leak detection; gaining the experimence and development of the ways to eliminate the consequences of accident caused by big water leak into sodium; accumulation of knowledge on restoration of MSG operating ability after accident; experimental test of calculational techniques for big leak accidents to use them in future for calculational studies of similar situations at other reactors equipped with sodium-water steam generators; refinement of characteristics of hydrodynamic and thermal effects interaction zone for big leak in real circuit during the plant operation. A series of experiments with the imitation of water leak into sodium by means of argon and steam supply through injection devices, located before the steam superheater module of one of the sections and between evaporator module of the same section, is conducted. The range of steam flow rate is 0.02-0.45 g/s. Duration of steam supply is 100-400 s. A conclusion is made that the results obtained can be used for steam generator of the BN-350 reactor [ru

  15. Fast reactor steam generators with sodium on the tube side. Design and operational parameters

    International Nuclear Information System (INIS)

    1994-01-01

    A comparison of design and operational characteristics as well as analysis of experience gained during the long terms operation of the Micro Module Inverse Steam Generator and Module Inverse Steam Generator at BOR 60 reactor are main aims of this technical report. 20 refs, 47 figs, 14 tabs

  16. 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

  17. Design study on steam generator integration into the VVER reactor pressure vessel

    International Nuclear Information System (INIS)

    Hort, J.; Matal, O.

    2004-01-01

    The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

  18. Systems design of direct-cycle supercritical-water-cooled fast reactors

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

    1995-01-01

    The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

  19. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  20. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  1. Sea water take-up facility for cooling reactor auxiliary

    International Nuclear Information System (INIS)

    Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

    1997-01-01

    The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

  2. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

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

  5. 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.

  6. 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

  7. Thermal-Hydraulic Analyses of Transients in an Actinide-Burner Reactor Cooled by Forced Convection of Lead Bismuth

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.

  8. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

    2008-03-15

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

  9. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    International Nuclear Information System (INIS)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R.

    2008-03-01

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to the

  10. Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Zhang Zuoyi; Dong Yujie; Scherer, Winfried

    2005-01-01

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

  11. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

  12. Lead- or Lead-bismuth-cooled fast reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Courouau, J.L.; Dufour, P.; Guidez, J.; Latge, C.; Martinelli, L.; Renault, C.; Rimpault, G.

    2014-01-01

    Lead-cooled fast reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. So far no lead-cooled reactors have existed in the world except lead-bismuth-cooled reactors in soviet submarines. Some problems linked to the use of the lead-bismuth eutectic appeared but were satisfactorily solved by a more rigorous monitoring of the chemistry of the lead-bismuth coolant. Lead presents various advantages as a coolant: no reactivity with water and the air,a high boiling temperature and low contamination when irradiated. The main asset of the lead-bismuth alloy is the drop of the fusion temperature from 327 C degrees to 125 C degrees. The main drawback of using lead (or lead-bismuth) is its high corrosiveness with metals like iron, chromium and nickel. The high corrosiveness of the coolant implies low flow velocities which means a bigger core and consequently a bigger reactor containment. Different research programs in the world (in Europe, Russia and the USA) are reviewed in the article but it appears that the development of this type of reactor requires technological breakthroughs concerning materials and the resistance to corrosion. Furthermore the concept of lead-cooled reactors seems to be associated to a range of low output power because of the compromise between the size of the reactor and its resistance to earthquakes. (A.C.)

  13. 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

  14. Passive cooling in modern nuclear reactors

    International Nuclear Information System (INIS)

    Rouai, N. M.

    1998-01-01

    This paper presents some recent experimental results performed with the aim of understanding the mechanism of passive cooling. The AP 600 passive containment cooling system is simulated by an electrically heated vertical pipe, which is cooled by a naturally induced air flow and by a water film descending under gravity. The results demonstrate that although the presence of the water film improved the heat transfer significantly, the mode of heat transfer was very dependent on the experimental parameters. Preheating the water improved both film stability and overall cooling performance

  15. SURGTANK, Steam Pressure, Saturation Temperature or Reactor Surge Tank

    International Nuclear Information System (INIS)

    Gorman, D.J.; Gupta, R.K.

    2001-01-01

    1 - Description of problem or function: SURGTANK generates the steam pressure, saturation temperature, and ambient temperature history for a nuclear reactor steam surge tank (pressurizer) in a state of thermodynamic equilibrium subjected to a liquid insurge described by a specified time history of liquid levels. It is capable also of providing the pressure and saturation temperature history, starting from thermodynamic equilibrium conditions, for the same tank subjected to an out-surge described by a time history of liquid levels. Both operations are available for light- or heavy- water nuclear reactor systems. The tank is assumed to have perfect thermal insulation on its outer wall surfaces. 2 - Method of solution: Surge tank geometry and initial liquid level and saturation pressure are provided as input for the out-surge problem, along with the prescribed time-sequence level history. SURGTANK assumes a reduced pressure for the end of the first change in liquid level and determines the associated change of entropy for the closed system. The assumed pressure is adjusted and the associated change in entropy recalculated until a pressure is attained for which no change occurs. This pressure is recorded and used as the beginning pressure for the next level increment. The system is then re-defined to exclude the small amount of liquid which has left the tank, and a solution for the pressure at the end of the second level increment is obtained. The procedure is terminated when the pressure at the end of the final increment has been determined. Surge tank geometry, thermal conductivity, specific heat, and density of tank walls, initial liquid level, and saturation pressure are provided as input for the insurge problem, along with the prescribed time-sequence level history. SURGTANK assumes a slightly in- creased pressure for the end of the first level, the inner tank sur- face is assumed to follow saturation temperature, linearly with time, throughout the interval, and

  16. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  17. Sorption-enhanced steam methane reforming in fluidized bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Johnsen, Kim

    2006-10-15

    Hydrogen is considered to be an important potential energy carrier; however, its advantages are unlikely to be realized unless efficient means can be found to produce it without generation of CO{sub 2}. Sorption-enhanced steam methane reforming (SE-SMR) represent a novel, energy-efficient hydrogen production route with in situ CO{sub 2} capture, shifting the reforming and water gas shift reactions beyond their conventional thermodynamic limits. The use of fluidized bed reactors for SE-SMR has been investigated. Arctic dolomite, a calcium-based natural sorbent, was chosen as the primary CO{sub 2}-acceptor in this study due to high absorption capacity, relatively high reaction rate and low cost. An experimental investigation was conducted in a bubbling fluidized bed reactor of diameter 0.1 m, which was operated cyclically and batch wise, alternating between reforming/carbonation conditions and higher-temperature calcination conditions. Hydrogen concentrations of >98 mole% on a dry basis were reached at 600 C and 1 atm, for superficial gas velocities in the range of {approx}0.03-0.1 m/s. Multiple reforming-regeneration cycles showed that the hydrogen concentration remained at {approx}98 mole% after four cycles. The total production time was reduced with an increasing number of cycles due to loss of CO{sub 2}-uptake capacity of the dolomite, but the reaction rates of steam reforming and carbonation seemed to be unaffected for the conditions investigated. A modified shrinking core model was applied for deriving carbonation kinetics of Arctic dolomite, using experimental data from a novel thermo gravimetric reactor. An apparent activation energy of 32.6 kj/mole was found from parameter fitting, which is in good agreement with previous reported results. The derived rate expression was able to predict experimental conversion up to {approx}30% very well, whereas the prediction of higher conversion levels was poorer. However, the residence time of sorbent in a continuous

  18. Strain measurement on a compact nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Scaldaferri, Denis Henrique Bianchi; Gomes, Paulo de Tarso Vida; Mansur, Tanius Rodrigues; Pozzo, Renato del; Mola, Jairo

    2011-01-01

    This work presents the strain measurement procedures applied to a compact nuclear reactor steam generator, during a hydrostatic test, using strain gage technology. The test was divided in two steps: primary side test and secondary side test. In the primary side test twelve points for strain measurement using rectangular rosettes, three points (two external and one internal) for temperature measurement using special strain gages and one point for pressure measurement using a pressure transducer were monitored. In the secondary side test 18 points for strain measurement using rectangular rosettes, four points (two external and two internal) for temperature measurement using special strain gages and one point for pressure measurement using a pressure transducer were monitored. The measurement points on both internal and external pressurizer walls were established from pre-calculated stress distribution by means of numerical approach (finite elements modeling). Strain values using a quarter Wheatstone bridge circuit were obtained. Stress values, from experimental strain were determined, and to numerical calculation results were compared. (author)

  19. 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)

  20. Leak detector for a steam generator in FBR type reactors

    International Nuclear Information System (INIS)

    Miyaji, Nobuyoshi.

    1979-01-01

    Purpose: To facilitate maintenance for liquid leak detectors such as exchange of nickel membrane sensors during operation in a sodium-cooled fbr type reactor. Constitution: A pipeway capable of supplying a cover gas such as argon into the cylinder of a hydrogen detector containing a nickel membrane sensor is provided in a liquid leak detector constituting a part of a by-pass loop. The pipeway is also adapted to be evacuated. A pipeway and a small sodium tank for drain use are provided on the side of the by-pass loop near valves. Then, after closing the inlet and outlet valves to disconnect the by-pass loop from the sodium main pipeway, the cover gas is supplied to drive liquid sodium to the drain tank. After the drain of the liquid sodium, the sensor can be replaced. (Ikeda, J.)

  1. Methane-steam reforming by molten salt - membrane reactor using concentrated solar thermal energy

    International Nuclear Information System (INIS)

    Watanuki, K.; Nakajima, H.; Hasegawa, N.; Kaneko, H.; Tamaura, Y.

    2006-01-01

    By utilization of concentrated solar thermal energy for steam reforming of natural gas, which is an endothermic reaction, the chemical energy of natural gas can be up-graded. The chemical system for steam reforming of natural gas with concentrated solar thermal energy was studied to produce hydrogen by using the thermal storage with molten salt and the membrane reactor. The original steam reforming module with hydrogen permeable palladium membrane was developed and fabricated. Steam reforming of methane proceeded with the original module with palladium membrane below the decomposition temperature of molten salt (around 870 K). (authors)

  2. Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

    International Nuclear Information System (INIS)

    Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

    1979-01-01

    Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

  3. 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)

  4. 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.

  5. Device for preventing cooling water from flowing out of reactor

    International Nuclear Information System (INIS)

    Chinen, Masanori; Kotani, Koichi; Murase, Michio.

    1976-01-01

    Object: To provide emergency cooling system, which can prevent cooling water bearing radioactivity from flowing to the outside of the reactor at the time of breakage of feedwater pipe, thus eliminating the possibility of exposure of the fuel rod to provide high reliability and also reducing the possibility of causing radioactive pollution. Structure: The device for preventing cooling water from flowing out from the reactor features a jet nozzle inserted in a feedwater pipe adjacent to the inlet or outlet thereof immediately before the reactor container. The nozzle outlet is provided in the vicinity of the reactor wall and in a direction opposite to the direction of out-flow, and water supplied from a high pressure pump is jetted from it. (Nakamura, S.)

  6. Technical and economic studies of small reactors for supply of electricity and steam

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-02-01

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out for several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 365 MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule, and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating, or desalination

  7. Technical and economic studies of small reactors for supply of electricity and steam

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-01-01

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313 MW(t) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(t), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(t) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating or desalination

  8. Technical and economic studies of small reactors for supply of electricity and steam

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-01-01

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating or desalination. (author)

  9. Analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1976-09-01

    The aim of this analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors is to get results on the possible extent of blockages and the time necessary for growth which may be used for a safety evaluation. After an introduction where the thermohydraulic and physical/chemical aspects of the problems are considered, the causes for the local cooling disturbances and the phenomena arising with it are freated in more detail. (orig./TK) [de

  10. Radial Microchannel Reactor (RMR) used in Steam Reforming CH4

    Science.gov (United States)

    2013-05-13

    steam reforming natural gas for a wide variety of application from distributed energy production...into synthesis gas . Synthesis gas is used in the production of hydrogen , in GTL and other chemical processes. Steam reforming in an RMR was studied...technology has the potential to have a transformational reduction in cost and size of steam reforming natural gas for a wide variety of application

  11. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

  12. Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1985-11-01

    This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed

  13. Online monitoring of steam/water chemistry of a fast breeder test reactor

    International Nuclear Information System (INIS)

    Subramanian, K.G.; Suriyanarayanan, A.; Thirunavukarasu, N.; Naganathan, V.R.; Panigrahi, B.S.; Jambunathan, D.

    2005-01-01

    Operating experience with the once-through steam generator of a fast breeder test reactor (FBTR) has shown that an efficient water chemistry control played a major role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. In order to meet the stringent feedwater and steam quality specifications, use of fast and sensitive online monitors to detect impurity levels is highly desirable. Online monitoring techniques have helped in achieving feedwater of an exceptional degree of purity. Experience in operating the online monitors in the steam/water system of a FBTR is discussed in detail in this paper. In addition, the effect of excess hydrazine in the feedwater on the steam generator leak detection system and the need for a hydrazine online meter are also discussed. (orig.)

  14. IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2011-01-01

    The IAEA, within the framework of its Nuclear Energy Department’s Technical Working Group on Fast Reactors (TWG-FR), assists Member States activities in fast reactors technology development areas by providing an umbrella for information exchange [topical Technical Meetings (TMs), Workshops and large Conferences] and collaborative R&D [Coordinated Research Projects (CRPs)]. The Technical meeting on “Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors” was held from 21 – 22 December 2011 in Vienna, addressing Member States’ expressed needs of information exchange in the field of advanced fast reactor design features, with particular attention to innovative heat exchangers and steam generators. The Objective of the TM is to provide a global forum for in-depth information exchange and discussion on the most advanced concepts of heat exchangers and steam generators for fast reactors. More specifically, the objectives are: · Review of the status of advanced fast reactor development activities with special emphasis on design and performance of heat exchangers and steam generators; · Discuss requirements for innovative heat exchangers and steam generators; · Present results of studies and conceptual designs for innovative heat exchangers and steam generators; · Provide recommendations for international collaboration under the IAEA aegis. The meeting agenda of the meeting is in Annex I

  15. Safety analysis of reactor's cooling system

    International Nuclear Information System (INIS)

    1999-01-01

    Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

  16. Some methods of failed fuel element detection in water cooled reactors

    International Nuclear Information System (INIS)

    Strindehag, O.M.

    1976-01-01

    The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

  17. Membrane steam reforming of natural gas for hydrogen production by utilization of medium temperature nuclear reactor

    International Nuclear Information System (INIS)

    Djati Hoesen Salimy

    2010-01-01

    The assessment of steam reforming process with membrane reactor for hydrogen production by utilizing of medium temperature nuclear reactor has been carried out. Difference with the conventional process of natural gas steam reforming that operates at high temperature (800-1000°C), the process with membrane reactor operates at lower temperature (~500°C). This condition is possible because the use of perm-selective membrane that separate product simultaneously in reactor, drive the optimum conversion at the lower temperature. Besides that, membrane reactor also acts the role of separation unit, so the plant will be more compact. From the point of nuclear heat utilization, the low temperature of process opens the chance of medium temperature nuclear reactor utilization as heat source. Couple the medium temperature nuclear reactor with the process give the advantage from the point of saving fossil fuel that give direct implication of decreasing green house gas emission. (author)

  18. Heat transfer and friction characteristics in steam cooled rectangular channels with rib turbulators

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Jianying; Gao, Tieyu; Li, Guojun [Xi' an Jiaotong University, Xi' an (China)

    2014-01-15

    We studied the heat transfer and friction characteristics in steam-cooled rectangular channels with rib turbulators on W side or H side walls in the Reynolds number (Re) range of 10000-80000. Each of the test channels was welded by four stainless steel plates to simulate the actual geometry and heat transfer structure of blade/vane internal cooling passage. The length of the channel L was 1000 mm, the cross section of the channel was 40 mm X 80 mm, and the pitch-to-rib height ratio p/e was kept at 10. The channel blockage ratio (W/H) was 0.047. Results showed that the Nusselt number (Nu) distributions displayed different trends at the entrance region with the increase of Re for the rib turbulators on the W side walls. The heat transfer performance of the rib turbulators on the H side walls was about 24- 27% higher than that on the W side walls at the same pumping power. In addition, semi-empirical correlations for the two cases, rib turbulators on W side walls and rib turbulators on H side walls, were developed based on the heat transfer results, which could be used in the design of the internal cooling passage of new generation steam-cooled gas turbine blade/vane.

  19. Status of liquid metal cooled fast reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-04-01

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants Refs, figs, tabs

  20. Conceptual design study of small lead-bismuth cooled reactor

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Hori, Toru; Kida, Masanori; Konomura, Mamoru

    2004-11-01

    In phase 2 of the feasibility study of commercialized fast reactor cycle systems of JNC, we make a concept of a small sodium cooled reactor for a power source of a city with various requirements, such as, safety and economical competitiveness. various reactor concepts are surveyed and a tank type reactor whose intermediate heat exchanger and primary main pumps are arranged in series is selected. In this study, a compact long life core and a simple reactor structure designs are pursued. The core type is three regional Zr concentration with one Pu enrichment core, the reactor outlet temperature achieves 550degC and the reactor electric output increases from 150 MWe to 165 MWe. The construction cost is much higher than the economical goal in the case of FOAK. But the construction cost in the case of NOAK is estimated to be 85.6% achieving the economical goal. (author)

  1. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    1999-04-01

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants

  2. A study of the external cooling capability for the prevention of reactor vessel failure

    Energy Technology Data Exchange (ETDEWEB)

    Chang, S H; Baek, W P; Moon, S K; Yang, S H; Kim, S H [Korea Advanced Institute of Science Technology, Daejeon (Korea, Republic of)

    1994-07-15

    This study (a 3-year program) aims to perform a comprehensive assessment of the feasibility of external vessel flooding with respect to advanced pressurized water reactor plants to be built in Korea. During the first year, review of the relevant phenomena and preliminary assessment of the concept have been performed. Also performed is a review of heat transfer correlations for the computer program that will be developed for assessment of the cooling capability of external vessel flooding. Important phenomena that determine the cooling capability of external vessel flooding are (a) the initial transient before formation of molten corium pool, (b) natural convection of in-vessel molten corium pool, (c) radiative heat exchange between the molten corium pool and the upper vessel structures, (d) thermal hydraulics outside the vessel, (e) structural integrity consideration, and (f) long-term phenomena. The adoption of the concept should be decided by considering several factors such as (a) vessel submergence procedure, (b) cooling requirements, (c) vessel design features, (d) steam production, (e) instrumentation needs, and (f) an overall accident management strategy. The external vessel cooling concept looks to be promising. However, further study is required for a reliable decision making. Several correlations are available for the prediction of cooling capability of the present concept. However, it is difficult to define a sufficiently reliable set of correlations; sensitivity studies would be required in assessing the cooling capability with the computer program.

  3. Method of cooling a pressure tube type reactor

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro.

    1983-01-01

    Purpose: To improve the operation efficiency of a nuclear reactor by carrying out cooling depending on the power distribution in the reactor core. Constitution: Reactor core channels are divided into a plurality of channel groups depending on the reactor power, and a water drum and a pump are disposed to each of the channel groups so as to increase the amount of coolants in response to the magnitude of the power from each of the channel groups. In this way, the minimum limiting power ratio can be increased. (Seki, T.)

  4. Analysis of steam explosions in plate-type, uranium-aluminum fuel test reactors

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1989-01-01

    The concern over steam explosions in nuclear reactors can be traced to prompt critical nuclear excursions in aluminum-clad/fueled test reactors, as well as to explosive events in aluminum, pulp, and paper industries. The Reactor Safety Study prompted an extensive analytical and experimental effort for over a decade. This has led to significant improvements in their understanding of the steam explosion issue for commercial light water reactors. However, little progress has been made toward applying the lessons learned from this effort to the understanding and modeling of steam explosion phenomena in aluminum-clad/fueled research and test reactors. The purposes of this paper are to (a) provide a preliminary analysis of the destructive events in test reactors, based on current understandings of steam explosions; (b) provide a proposed approach for determining the likelihood of a steam explosion event under scenarios in which molten U-Al fuel drops into a water-filled cavity; and (c) present a benchmarking study conducted to estimate peak pressure pulse magnitudes

  5. Pd-Ag membrane reactor for steam reforming reactions: a comparison between different fuels

    NARCIS (Netherlands)

    Gallucci, F.; Basile, A.

    2008-01-01

    The simulation of a dense Pd-based membrane reactor for carrying out the methane, the methanol and the ethanol steam reforming (SR) reactions for pure hydrogen production is performed. The same simulation is also performed in a traditional reactor. This modelling work shows that the use of membrane

  6. A dense Pd/Ag membrane reactor for methanol steam reforming: Experimental study

    NARCIS (Netherlands)

    Basile, A.; Gallucci, F.; Paturzo, L.

    2005-01-01

    This paper focuses on an experimental study of the methanol steam reforming (MSR) reaction. A dense Pd/Ag membrane reactor (MR) has been used, and its behaviour has been compared to the performance of a traditional reactor (TR) packed with the same catalyst type and amount. The parameters

  7. 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

  8. A modular approach to lead-cooled reactors modelling

    Energy Technology Data Exchange (ETDEWEB)

    Casamassima, V. [CESI RICERCA, via Rubattino 54, I-20134 Milano (Italy)], E-mail: casamassima@cesiricerca.it; Guagliardi, A. [CESI RICERCA, via Rubattino 54, I-20134 Milano (Italy)], E-mail: guagliardi@cesiricerca.it

    2008-06-15

    After an overview of the lego plant simulation tools (LegoPST), the paper gives some details about the ongoing LegoPST extension for modelling lead fast reactor plants. It refers to a simple mathematical model of the liquid lead channel dynamic process and shows the preliminary results of its application in dynamic simulation of the BREST 300 liquid lead steam generator. Steady state results agree with reference data [IAEA-TECDOC 1531, Fast Reactor Database, 2006 Update] both for water and lead.

  9. A modular approach to lead-cooled reactors modelling

    International Nuclear Information System (INIS)

    Casamassima, V.; Guagliardi, A.

    2008-01-01

    After an overview of the lego plant simulation tools (LegoPST), the paper gives some details about the ongoing LegoPST extension for modelling lead fast reactor plants. It refers to a simple mathematical model of the liquid lead channel dynamic process and shows the preliminary results of its application in dynamic simulation of the BREST 300 liquid lead steam generator. Steady state results agree with reference data [IAEA-TECDOC 1531, Fast Reactor Database, 2006 Update] both for water and lead

  10. A study on different thermodynamic cycle schemes coupled with a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Qu, Xinhe; Yang, Xiaoyong; Wang, Jie

    2017-01-01

    Highlights: • The features of three different power generation schemes, including closed Brayton cycle, non-reheating combined cycle and reheating combined cycle, coupled with high temperature gas-cooled reactor (HTGR) were investigated and compared. • The effects and mechanism of reactor core outlet temperature, compression ratio and other key parameters over cycle characteristics were analyzed by the thermodynamic models.. • It is found that reheated combined cycle has the highest efficiency. Reactor outlet temperature and main steam parameters are key factors to improve the cycle’s performance. - Abstract: With gradual increase in reactor outlet temperature, the efficient power conversion technology has become one of developing trends of (very) high temperature gas-cooled reactors (HTGRs). In this paper, different cycle power generation schemes for HTGRs were systematically studied. Physical and mathematical models were established for these three cycle schemes: closed Brayton cycle, simple combined cycle, and reheated combined cycle. The effects and mechanism of key parameters such as reactor core outlet temperature, reactor core inlet temperature and compression ratio on the features of these cycles were analyzed. Then, optimization results were given with engineering restrictive conditions, including pinch point temperature differences. Results revealed that within the temperature range of HTGRs (700–900 °C), the reheated combined cycle had the highest efficiency, while the simple combined cycle had the lowest efficiency (900 °C). The efficiencies of the closed Brayton cycle, simple combined cycle and reheated combined cycle are 49.5%, 46.6% and 50.1%, respectively. These results provide insights on the different schemes of these cycles, and reveal the effects of key parameters on performance of these cycles. It could be helpful to understand and develop a combined cycle coupled with a high temperature reactor in the future.

  11. 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)

  12. 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.

  13. Design study on sodium-cooled large-scale reactor

    International Nuclear Information System (INIS)

    Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

    2002-05-01

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2001, which is the first year of Phase 2. In the JFY2001 design study, a plant concept has been constructed based on the design of the advanced loop type reactor, and fundamental specifications of main systems and components have been set. Furthermore, critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  14. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  15. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  16. 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

  17. Decay heat removal analyses in heavy-liquid-metal-cooled fast breeding reactors. Development of the thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Takaaki; Enuma, Yasuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwasaki, Takashi [Nuclear Energy System Inc., Tokyo (Japan); Ohyama, Kazuhiro [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2001-05-01

    The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with all the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300 - 550 MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400 MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional Steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100 MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxiliary Cooling System (PRACS) in the 400 MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate thermal-hydraulic phenomena during steady-state and transient conditions. (author)

  18. 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

  19. A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Laboratory; Zhang, Hongbin [Idaho National Laboratory; Zou, Ling [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2015-03-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety

  20. The atmospheric cooling of nuclear power stations

    International Nuclear Information System (INIS)

    Leuenberger, J.M.; Mayor, J.C.; Gassmann, F.; Lieber, K.

    1978-08-01

    Four different types of nuclear reactor are considered: light water reactors, high temperature reactors with steam circulation and with direct gas turbine circulation, and fast breeder reactors. Wet and dry cooling towers are described and experimental studies carried out using cooling tower models are presented. (G.T.H.)

  1. The performance of a temperature cascaded cogeneration system producing steam, cooling and dehumidification

    KAUST Repository

    Myat, Aung

    2013-02-01

    This paper discusses the performance of a temperature-cascaded cogeneration plant (TCCP), equipped with an efficient waste heat recovery system. The TCCP, also called a cogeneration system, produces four types of useful energy-namely, (i) electricity, (ii) steam, (iii) cooling and (iv) dehumidification-by utilizing single fuel source. The TCCP comprises a Capstone C-30 micro-turbine that generates nominal capacity of 26 kW of electricity, a compact and efficient waste heat recovery system and a host of waste-heat-activated devices, namely (i) a steam generator, (ii) an absorption chiller, (iii) an adsorption chiller and (iv) a multi-bed desiccant dehumidifier. The performance analysis was conducted under different operation conditions such as different exhaust gas temperatures. It was observed that energy utilization factor could be as high as 70% while fuel energy saving ratio was found to be 28%. © 2013 Desalination Publications.

  2. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  3. Passive safety features in current and future water cooled reactors

    International Nuclear Information System (INIS)

    1990-11-01

    Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

  4. Appraisal of possible combustion hazards associated with a high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Palmer, H.B.; Sibulkin, M.; Strehlow, R.A.; Yang, C.H.

    1978-03-01

    The report presents a study of combustion hazards that may be associated with the High Temperature Gas Cooled Reactor (HTGR) in the event of a primary coolant circuit depressurization followed by water or air ingress into the prestressed concrete reactor vessel (PCRV). Reactions between graphite and steam or air produce the combustible gases H 2 and/or CO. When these gases are mixed with air in the containment vessel (CV), flammable mixtures may be formed. Various modes of combustion including diffusion or premixed flames and possibly detonation may be exhibited by these mixtures. These combustion processes may create high over-pressure, pressure waves, and very hot gases within the CV and hence may threaten the structural integrity of the CV or damage the instrumentation and control system installations within it. Possible circumstances leading to these hazards and the physical characteristics related to them are delineated and studied in the report

  5. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

    2016-01-01

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

  6. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  7. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    Science.gov (United States)

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  8. 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

  9. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.

    1997-01-01

    This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs

  10. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

    2013-01-01

    • Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

  11. Method of operating a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Lysell, G.

    1975-01-01

    When operating a water-cooled nuclear reactor, in which the fuel rods consist of zirconium alloy tubes containing an oxidic nuclear fuel, stress corrosion in the tubes can be reduced or avoided if the power of the reactor is temporarily increased so much that the thermal expansion of the nuclear fuel produces a flow of the material in the tube. After that temporary power increase the power output is reduced to the normal power

  12. 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.

  13. Analysis of a main steam isolation value closure anticipated transient without scram in a boiling water reactor

    International Nuclear Information System (INIS)

    Liaw, T.J.; Pan, C.; Chen, G.S.

    1989-01-01

    Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main steam isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4. Without boron injection and makeup coolant, the reactor loses its coolant inventory very quickly and the reactor power drops rapidly to ∼ 16% of rated power due to negative void reactivity. With coolant makeup from the high-pressure core spray and the reactor core isolation cooling systems, the rector reaches a quasi-steady-state condition after an initially rapidly changing transient. The dome pressure, downcomer water level, and core power oscillate around a mean value; the average core power is ∼ 15%, which is approximately equal to the power needed to heat and evaporate the subcooled makeup coolant. Lower boron concentrations in the core tend to complicate reactor behavior due to the combination of two competing phenomena: the negative boron reactivity and the positive reactivity caused by a void collapse

  14. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    International Nuclear Information System (INIS)

    Woo, H.H.; Lu, S.C.

    1981-01-01

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

  15. Temperature monitoring of gas-cooled reactors

    International Nuclear Information System (INIS)

    Kaiser, G.E.

    1977-01-01

    The present paper deals with questions like : a) Why temperature monitoring in high-temperature reactors at all. b) How are the measuring positions arranged and how are the measurements designed. c) What technique of temperature measurement is applied. (RW) [de

  16. Capital cost: gas cooled fast reactor plant

    International Nuclear Information System (INIS)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design

  17. Emergency core cooling system in BWR type reactors

    International Nuclear Information System (INIS)

    Takizawa, Yoji

    1981-01-01

    Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

  18. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Boldyrev, V.M.; Mikhaj, V.I.

    1985-01-01

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  19. Replacement of the cooling system of the TRIGA Mainz reactor

    International Nuclear Information System (INIS)

    Menke, H.

    1988-01-01

    The inspection of the reactor facility resulted in a recommendation to install a new heat exchanger and at the same time to separate the primary cooling circuit and the water purification system. Due to possible the deposition of lime and organic matter on the tubes, the heat transfer rate has decreased. In the meantime a rule has been introduced, according to which the pressure in the secondary cooling circuit must be permanently higher than in the primary cooling circuit which prompted the design of a new cooling system. The detail planning was completed in December 1987. In response to the regulatory requirements a motion for a replacement of the cooling system was submitted to the authorities. The start of the procedure is possible a year after the obtaining of the licenses. In the planning of the changes an upgrading of the steady state power to 300 kW is envisioned

  20. Nuclear reactor, its cooling facility, nuclear power plant, and method of operating the same

    International Nuclear Information System (INIS)

    Tate, Hitoshi; Tominaga, Kenji; Fujii, Tadashi.

    1993-01-01

    The upper surface of inner structural materials in a container is partitioned by concrete structural walls to form an upper space portion. A pressure relief plate is disposed on the concrete structural walls. If an accident occurs, the pressure relief plate is operated to form a circulation path for a gas to return to the upper space portion again from the upper space portion. The temperature of cooling water in a pressure suppression chamber, that is, a wet well liquid phase portion is elevated by after heat of a reactor core. Evaporated steams transfer from the wet well gas phase portion to the upper space portion passing through pipelines and are mixed with N 2 gas present in the upper space portion. The mixed gas is cooled by a container inner wall cooled by air passing through an air cooling duct, flows downward by way of the pressure relief plate and reaches the wet well gas phase portion again. Since the gases in the upper space circulate by a driving force caused by the after heat, reliability of cooling performance can be improved upon occurrence of an accident without using an active driving force. (I.N.)

  1. Substantiation of vibration strength of nuclear reactor and steam generator internals. Main problems

    International Nuclear Information System (INIS)

    Fyodorov, V.G.; Sinyavasky, V.F.

    1977-01-01

    The report details the scope and priority of studies necessary for substantiation of vibration strength of steam generator tube bundles and reactor fuel assemblies, and design modifications helping to reduce flow-induced vibration of the internals specified. Steam generator tube bundles are studied on the basis of a standard establishing vibration requirements at various stages of design, manufacture and operation of a steam generator at a nuclear power station. The main vibration characteristics of tubes obtained through model and full-scale tests are compared with calculation results. Results are provided concerning test-stand vibration tests of fuel elements and fuel assemblies. (author)

  2. Heat exchanging tube behaviour in steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Pastor, D.; Oertel, K.

    1979-01-01

    Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

  3. Hydrogen production by biomass steam gasification in fluidized bed reactor with Co catalyst

    International Nuclear Information System (INIS)

    Kazuhiko Tasaka; Atsushi Tsutsumi; Takeshi Furusawa

    2006-01-01

    The catalytic performances of Co/MgO catalysts were investigated in steam gasification of cellulose and steam reforming of tar derived from cellulose gasification. For steam reforming of cellulose tar in a secondary fixed bed reactor, 12 wt.% Co/MgO catalyst attained more than 80% of tar reduction. The amount of produced H 2 and CO 2 increased with the presence of catalyst, and kept same level during 2 hr at 873 K. It is indicated that steam reforming of cellulose tar proceeds sufficiently over Co/MgO catalyst. For steam gasification of cellulose in a fluidized bed reactor, it was found that tar reduction increases with Co loading amount and 36 wt.% Co/MgO catalyst showed 84% of tar reduction. The amounts of produced gas kept for 2 hr indicating that 36 wt.% Co/MgO catalyst is stable during the reaction. It was concluded that these Co catalysts are promising systems for the steam gasification of cellulose and steam reforming of cellulose tar. (authors)

  4. Detection of steam leaks into sodium in fast reactor steam generators by acoustic techniques - An overview of Indian programme

    International Nuclear Information System (INIS)

    Prabhakar, R.; Vyjayanthi, R.K.; Kale, R.D.

    1990-01-01

    Realising the potential of acoustic leak detection technique, an experimental programme was initiated a few years back at Indira Gandhi Centre for Atomic Research (IGCAR) to develop this technique. The first phase of this programme consists of experiments to measure background noise characteristics on the steam generator modules of the 40 MW (thermal) Fast Breeder Test Reactor (FBTR) at Kalpakkam and experiments to establish leak noise characteristics with the help of a leak simulation set up. By subjecting the measured data from these experiments to signal analysis techniques, a criterion for acoustic leak detection for FBTR steam generator will be evolved. Second phase of this programme will be devoted to developing an acoustic leak detection system suitable for installation in the 500 MWe Prototype Fast Breeder Reactor (PFBR). This paper discusses the first phase of the experimental programme, results obtained from measurements carried out on FBTR steam generators and results obtained from leak simulation experiments. Acoustic leak detection system being considered for PFBR is also briefly described. 4 refs, 8 figs, 1 tab

  5. Chemical operational experience with the water/steam-circuit at KNK II; Presentation at the meeting on Experience exchange on operational experience of fast breeder reactors, Karlsruhe/Bensberg/Kalkar, June 18. - 22. 1990

    International Nuclear Information System (INIS)

    Grumer, U.

    1990-06-01

    The availability of sodium cooled reactors depends essentially from the safety and reliability of the sodium heated steam generator. The transition from experimental plants with 12-20 MW electrical power to larger plants with 600 MW (BN-600) or 1200 MW (Superphenix) required the change from modular components to larger and compact steam generators with up to 800 MW. Defects of these large components cause extreme losses in availability of the plant and have to be avoided. In view of this request, a comprehensive test program has been performed at KNK II in addition to the normal control of the water/steam-circuit to compile all operational data on the water and steam side of the sodium heated steam generator. This paper describes the plant and the water/steam-circuit with its mode of operation. The experience with the surveillance and different methods of the conditioning are discussed in detail in this presentation

  6. Indian programme on molten salt cooled nuclear reactors

    International Nuclear Information System (INIS)

    DuIera, I.V.; Vijayan, P.K.; Sinha, R.K.

    2013-01-01

    Bhabha Atomic Research Centre (BARC) is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of molten fluoride salts and is capable of supplying process heat at 1000 ℃ to facilitate hydrogen production by splitting water. BARC has also initiated studies for a reactor concept in which salts of molten fluoride fuel and coolant in fluid form, flows through the reactor core of graphite moderator, resulting in nuclear fission within the molten salt. For thorium fuel cycle, this concept is very attractive, since the fuel can be re-processed on-line, enabling it to be an efficient neutron breeder. (author)

  7. Ultrasonic sweep arm for sodium cooled reactors

    International Nuclear Information System (INIS)

    Rohrbacher, H.A.; Bartholomay, R.

    1975-05-01

    This report describes experience in the use of a new type of monitoring and testing device to be applied in conjunction with components under sodium. In the method outlined, ultrasonic pulses are used which are emitted into the sodium plenum of fast breeder reactors by newly developed high temperature transducers. The basic work was conducted under out-of-pile conditions in a sodium tank of the sodium tank facility of the Karlsruhe Institute for Reactor Development. The sensor development, which preceded this phase, resulted in the use of soldered lithium niobate crystals whose operating characteristics were improved by the preliminary treatment outlined in the report. Special materials and techniques suitable for sensor fabrication are proposed. An alternative to soldering is suggested for contacting the crystals with their diaphragms, i.e. a contact pressure concept for the range of application up to 2 MHz. (orig.) [de

  8. Gamma spectroscopy in water cooled reactors

    International Nuclear Information System (INIS)

    Persault, M.

    1977-10-01

    Gamma spectroscopy analysis of spent fuels in power reactors; study of two typical cases: determination of the power distribution by the mean of the activity of a low periodic element (Lanthanum 140) and determination of the burnup absolute rate by examining the ratio of Cesium 134 and Cesium 137 activities. Measures were realized on fuel solutions and on fuel assemblies. Development of a power distribution map of the assemblies and comparison with the results of a three dimensional calculation of core evolution [fr

  9. Investigation of thermodynamic cycle for generic 1200 MW{sub el} pressure channel reactor with nuclear steam superheat

    Energy Technology Data Exchange (ETDEWEB)

    Vincze, A.; Sidawi, K.; Abdullah, R.; Baldock, M.; Saltanov, E.; Pioro, I., E-mail: andrei.vincze@uoit.net, E-mail: khalil.sidawi@uoit.net, E-mail: rand.abdullah@uoit.net, E-mail: matthew.baldock@uoit.net, E-mail: eugene.saltanov@uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

    2014-07-01

    Current Nuclear Power Plants (NPPs) play a significant role in energy production around the world. All NPPs operating today employ a Rankine steam cycle for the conversion of thermal power to electricity. This paper will examine the steam cycle arrangement an experimental pressure channel reactor using Nuclear Steam Superheat (NSS) and compare it to two advanced reactor designs, the Advanced CANDU Reactor 1000 (ACR-1000) and the Advanced Boiling Water Reactor (ABWR) designs. The thermodynamic cycle layout and thermal efficiencies of the three reactor types will be discussed. (author)

  10. Primary cooling system for BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

    1993-01-01

    The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

  11. Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

    International Nuclear Information System (INIS)

    Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

    2008-01-01

    In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

  12. Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

    Science.gov (United States)

    Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

    2012-06-01

    Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

  13. 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)

  14. Modeling of heat transfer in wall-cooled tubular reactors

    NARCIS (Netherlands)

    Koning, G.W.; Westerterp, K.R.

    1999-01-01

    In a pilot scale wall-cooled tubular reactor, temperature profiles have been measured with and without reaction. As a model reaction oxidation of carbon monoxide in air over a copper chromite catalyst has been used. The kinetics of this reaction have been determined separately in two kinetic

  15. 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

  16. 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.

  17. 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

  18. Numerical modeling of a nuclear production reactor cooling lake

    International Nuclear Information System (INIS)

    Hamm, L.L.; Pepper, D.W.

    1987-01-01

    A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation

  19. Procedure for operating a heavy water cooled power reactor

    International Nuclear Information System (INIS)

    Rau, P.; Kumpf, H.

    1981-01-01

    Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

  20. Improvement of steam separator in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, Jan Peter; Cremer, Ingo; Lorenz, Maik [AREVA GmbH, Erlangen (Germany)

    2013-07-01

    The potential to improve the function of the steam separator is identified and explored by scaled air-water tests and validated CFD calculations. It can be outlined that requirements on a modern steam separator for BWR plants will be fulfilled, combined with very good operational experience of the existing separator designs (e.g. material, layout). With the new steam separator design, modern high performance fuel assembly designs can be used for various core loading strategies (e.g. low leakage). This allows an increased thermal power of up to +50 % for the fuel element clusters in the center of the core with high radial peaking factors. In addition, any problems with unallowable high moisture at the turbine are solved with the new design, which have been identified for running BWR plants with the old steam separator design after applying new core loading patterns (e.g. after power uprates). A compatible steam separator design for all running BWRs is ready to launch. (orig.)

  1. A single-stage high pressure steam injector for next generation reactors: test results and analysis

    International Nuclear Information System (INIS)

    Cattadori, G.; Galbiati, L.; Mazzocchi, L.; Vanini, P.

    1995-01-01

    Steam injectors can be used in advanced light water reactors (ALWRs) for high pressure makeup water supply; this solution seems to be very attractive because of the ''passive'' features of steam injectors, that would take advantage of the available energy from primary steam without the introduction of any rotating machinery. The reference application considered in this work is a high pressure safety injection system for a BWR; a water flow rate of about 60 kg/s to be delivered against primary pressures covering a quite wide range up to 9 MPa is required. Nevertheless, steam driven water injectors with similar characteristics could be used to satisfy the high pressure core coolant makeup requirements of next generation PWRs. With regard to BWR application, an instrumented steam injector prototype with a flow rate scaling factor of about 1:6 has been built and tested. The tested steam injector operates at a constant inlet water pressure (about 0.2 MPa) and inlet water temperature ranging from 15 to 37 o C, with steam pressure ranging from 2.5 to 8.7 MPa, always fulfilling the discharge pressure target (10% higher than steam pressure). To achieve these results an original double-overflow flow rate-control/startup system has been developed. (Author)

  2. Reactor coolant pump seal response to loss of cooling

    International Nuclear Information System (INIS)

    Graham, T.; Metcalfe, R.; Burchett, P.

    2000-01-01

    This paper describes the results of a test done to determine the performance of a reactor coolant pump seal for a water cooled nuclear reactor under loss of all cooling conditions. Under these conditions, seal faces can lose their liquid lubricating film and elastomers can rapidly degrade. Temperatures in the seal-cartridge tester reached 230 o C in three hours, at which time the tester was stopped and the temperature increased to 265 o C for a further five hours before cooling was restored. Seal leakage was 'normal' throughout the test. Parts sustained minor damage with no effect on seal integrity. Plant operators were shown to have ample margin beyond their 15 minute allowable reaction time. (author)

  3. Reverse primary-side flow in steam generators during natural circulation cooling

    International Nuclear Information System (INIS)

    Stumpf, H.; Motley, F.; Schultz, R.; Chapman, J.; Kukita, Y.

    1987-01-01

    A TRAC model of the Large Scale Test Facility with a 3-tube steam-generator model was used to analyze natural-circulation test ST-NC-02. For the steady state at 100% primary mass inventory, TRAC was in excellent agreement with the natural-circulation flow rate, the temperature distribution in the steam-generator tubes, and the temperature drop from the hot leg to the steam-generator inlet plenum. TRAC also predicted reverse flow in the long tubes. At reduced primary mass inventories, TRAC predicted the three natural-circulation flow regimes: single phase, two phase, and reflux condensation. TRAC did not predict the cyclic fill-and-dump phenomenon seen briefly in the test. TRAC overpredicted the two-phase natural-circulation flow rate. Since the core is well cooled at this time, the result is conservative. An important result of the analysis is that TRAC was able to predict the core dryout and heatup at approximately the same primary mass inventory as in the test. 4 refs., 8 figs., 2 tabs

  4. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  5. Proceedings of 'workshop on Pb-alloy cooled fast reactor'

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Kim, Yong Hee; Hong, Ser Gi

    2003-06-01

    The objective of 'Workshop on Pb-Alloy Cooled Fast Reactor', held in Taejeon, Korea on May 6, 2003, is to enhance the basic knowledge in this area by facilitating the exchange of information and discussions about problematic area of design aspects. There were five presentations from three different countries and about 25 participants gathered during the workshop. The topics covered in the workshop include benefits and drawbacks of Pb-alloy and Sodium coolant, two Pb-alloy cooled 900 MWt reactor designs using both B4C rods and NSTs, BREST-300 breakeven reactor and transmutation effectiveness of LLFPs in the typical thermal/fast neutron systems. The generic conclusion for the Pb-alloy cooled fast reactor from this workshop is as follows: 1) It has a potential to satisfy the goals established for the Generation-IV reactor concepts, so it has a bright future. 2) As a fast neutron system with a moderate breeding or a conversion, it is flexible in its roles and has superior safety characteristics over sodium coolant because of Pb-alloy's chemical inertness with water/air and high boiling temperature

  6. 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

  7. Water box for steam generator

    International Nuclear Information System (INIS)

    Lecomte, Robert; Viaud, Michel.

    1975-01-01

    This invention relates to a water box for connecting an assembly composed of a vertical steam generator and a vertical pump to the vessel of the nuclear reactor, the assembly forming the primary cooling system of a pressurised water reactor. This invention makes it easy to dismantle the pump on the water box without significant loss of water in the primary cooling system of the reactor and particularly without it being necessary to drain the water contained in the steam generator beforehand. It makes it possible to shorten the time required for dismantling the primary pump in order to service or repair it and makes dismantling safer in that the dismantling does not involve draining the steam generator and therefore the critical storage of a large amount of cooling water that has been in contact with the fuel assemblies of the nuclear reactor core [fr

  8. Steam conversion of liquefied petroleum gas and methane in microchannel reactor

    Science.gov (United States)

    Dimov, S. V.; Gasenko, O. A.; Fokin, M. I.; Kuznetsov, V. V.

    2018-03-01

    This study presents experimental results of steam conversion of liquefied petroleum gas and methane in annular catalytic reactor - heat exchanger. The steam reforming was done on the Rh/Al2O3 nanocatalyst with the heat applied through the microchannel gap from the outer wall. Concentrations of the products of chemical reactions in the outlet gas mixture are measured at different temperatures of reactor. The range of channel wall temperatures at which the ratio of hydrogen and carbon oxide in the outlet mixture grows substantially is determined. Data on the composition of liquefied petroleum gas conversion products for the ratio S/C = 5 was received for different GHVS.

  9. 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

  10. Balancing passive and active systems for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Fil, N.S.; Allen, P.J.; Kirmse, R.E.; Kurihara, M.; Oh, S.J.; Sinha, R.K.

    1999-01-01

    Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

  11. Auxiliary equipment for cooling water in a reactor

    International Nuclear Information System (INIS)

    Konno, Yasuhiro; Sakairi, Toshiaki.

    1975-01-01

    Object: To effectively make use of pressure energy of reactor water, which has heretofore been discarded, to enable supply of emergency power supply of high reliability and to prevent spreading of environmental contamination. Structure: Sea water pumped by a sea water supply pump is fed to a heat exchanger. Reactor water carried through piping on the side to be cooled is removed in heat by the heat exchanger to be cooled and returned, and then again returned to the reactor. On the other hand, sea water heated by the heat exchanger is fed to a water wheel to drive the water wheel, after which it is discharged into a discharging path. A generator may be directly connected to the water wheel to use the electricity generated by the generator as the emergency power source. (Kamimura, M.)

  12. 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

  13. Research and development of the supercritical-pressure light water cooled reactor

    International Nuclear Information System (INIS)

    Oka, Yoshiaki

    2003-01-01

    The concept of high temperature reactor cooled by light water (SCR) has been developed at the University of Tokyo since 1989. Major elements of reactor conceptual design and safety were studied. It includes fuel rod design, core design of thermal and fast reactors, plant heat balance, safety design, accident and transient analysis, LOCA, PSA, plant control, start-up and stability. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical FPP in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil Fired Power Plants (FPP) will be fully utilized for SCR. Although the concept was developed at the University of Tokyo mostly with our own funds and resources, four funding was/is provided for the research in Japan so far. Those are TEPCO studies with Japanese vendors in 1994 and 1995. JSPS (Monbusho) funding of pulse radiolysis of supercritical water to the University of Tokyo, Japanese-NERI program of METI to Toshiba team on thermal hydraulics, corrosion and plant system and Japanese-NERI program of MEXT on water chemistry to the University of Tokyo. The concept was taken as the reference of HPLWR study in Europe with funding of EU in 2000 and 2001. The concept was evaluated in the Generation 4 reactor program in USA. It was selected as only one water-cooled Generation 4 reactor. This paper describes the overview of the conceptual design at the University of Tokyo and R and D in the world

  14. The United States fluoride-salt-cooled high-temperature reactor program

    International Nuclear Information System (INIS)

    Holcomb, David E.

    2013-01-01

    The United States is pursuing the development of fluoride-salt-cooled high-temperature reactors (FHRs) through the Department of Energy's Office of Nuclear Energy (DOE-NE). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. FHRs, in principle, have the potential to economically generate large amounts of electricity while maintaining full passive safety. FHRs, however, remain a longer-term power production option. A principal development focus is, thus, on shortening, to the extent possible, the overall development time by focusing initial efforts on the longest lead-time issues. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid-metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High-temperature gas-cooled reactors provide experience with coated-particle fuel and graphite components. Light-water reactors show the potential of transparent, high-heat-capacity coolants with low chemical reactivity. The FHR development efforts include both reactor concept and technology developments and are being broadly pursued. Oak Ridge National Laboratory (ORNL) provides technical leadership to the effort and is performing concept development on both a large base-load-type FHR as well as a small modular reactor (SMR) in addition to performing a broad scope of technology developments. Idaho National Laboratory (INL) is providing coated-particle fuel irradiation testing as well as developing high-temperature steam generator technology. The Massachusetts Institute of Technology (MIT

  15. 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)

  16. Natural-draught cooling tower of the Philippsburg-1 reactor

    International Nuclear Information System (INIS)

    Ernst, G.; Wurz, D.

    1983-01-01

    In spring 1980 a comprehensive research programm was carried out on the natural-draught cooling tower of the Philippsburg-1 reactor. The study was meant to synchronously acquire all parameters necessary for the evaluation of plant operation and cooling tower emissions. The study is subdivided into 8 sub-projects. Parts 1 to 7 that are included in this progress-of-work report describe experimental work and discuss the results. A critical analysis of measuring results proves that the values for operational behaviour and cooling tower emissions were duly anticipated. Even a very critical judgment of the results can exclude direct or indirect hazards for humans, animals and plants owing to cooling tower emissions. Sub-project 8 compares results from diffusion calculations (24 models) to results gained from experiments. The results of sub-project 8 will be published in a progress report to come. (orig.) [de

  17. Development of a helical-coil double wall tube steam generator for 4S reactor

    International Nuclear Information System (INIS)

    Kitajima, Yuko; Maruyama, Shigeki; Jimbo, Noboru; Hino, Takehisa; Sato, Katsuhiko

    2011-01-01

    The 4S, Super-Safe Small and Simple, is a small-sized sodium-cooled fast reactor. A fast reactor usually uses sodium as a coolant to transfer heat from core to turbine/generator system. The heat of the intermediate heat transport system and that of the water stream systems are exchanged by the steam generator (SG) tubes. If the tube failure occurs, a sodium/water reaction could be occurred. To prevent the reaction and enhance safety, a helical-coil-type double wall tube with wire mesh interlayer and continuous monitoring systems of tube failure are applied to the SG of the 4S. The development and general features of this type double wall tube were described in Ref. 1) and Ref. 2). Those paper summarized following results; The tubes studied in these references were straight type. To establish this SG, development of manufacturing method of helical-coil-type double wall tube and validation of the tube failure monitoring system are needed. In this study, three demonstration tests have been performed; welding test of the double wall tube to manufacture the tubes with 70-80m length, assembling test of the helical-coil tube, and confirmation test of the tube processing system using the fabricated helical-coil tubes. As a result, following technologies have been successfully established. (1) Development of the welding techniques for manufacturing of the helical-coil-type double wall tube with wire mesh interlayer. (2) The confirmation test for manufacturing the helical coil tube of the SG. (author)

  18. 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

  19. Effect of thermal barrier coatings on the performance of steam and water-cooled gas turbine/steam turbine combined cycle system

    Science.gov (United States)

    Nainiger, J. J.

    1978-01-01

    An analytical study was made of the performance of air, steam, and water-cooled gas-turbine/steam turbine combined-cycle systems with and without thermal-barrier coatings. For steam cooling, thermal barrier coatings permit an increase in the turbine inlet temperature from 1205 C (2200 F), resulting in an efficiency improvement of 1.9 percentage points. The maximum specific power improvement with thermal barriers is 32.4 percent, when the turbine inlet temperature is increased from 1425 C (2600 F) to 1675 C (3050 F) and the airfoil temperature is kept the same. For water cooling, the maximum efficiency improvement is 2.2 percentage points at a turbine inlet temperature of 1683 C (3062 F) and the maximum specific power improvement is 36.6 percent by increasing the turbine inlet temperature from 1425 C (2600 F) to 1730 C (3150 F) and keeping the airfoil temperatures the same. These improvements are greater than that obtained with combined cycles using air cooling at a turbine inlet temperature of 1205 C (2200 F). The large temperature differences across the thermal barriers at these high temperatures, however, indicate that thermal stresses may present obstacles to the use of coatings at high turbine inlet temperatures.

  20. Roof slab cooling device in a FBR type reactor

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1987-01-01

    Purpose: To obtain a roof slab cooling device capable of retaining cooling performance even in a case of electric power supply stop or failure and effective from economical point of view. Constitution: Atmospheric air is introduced into the cooling chamber of a proof slab and spontaneously passed to a exit pipeway connected to a stack thereby cooling the roof slab. Specifically, atmospheric air entered from the inlet pipeway is introduced to the cooling chamber and absorbs heat generate from the inside of the reactor container. Warmed air is sucked from the exit pipeway and then released into the atmosphere passing through the stack. The air cools the roof slab during circulation due to spontaneous passage and keeps the slab at a low temperature. Since the air is passed spontaneously, no power such as for a blower is required at all and, if the electric power supply should be lost, the cooling power can be maintained as it is to provide a high reliability. Further, since no electric power is required for the blowing power, it has high economical merit. (Horiuchi, T.)

  1. Corrosion by cooling gases in nuclear reactors

    International Nuclear Information System (INIS)

    Darras, R.

    1960-01-01

    This article begins with a review of the various materials which can be used and the cooling gases in which they may be heated, emphasis being placed on the importance of reaching temperatures as high as possible. This is followed by a few general remarks on the dry oxidation of metals and alloys, particularly with regard to diffusion phenomena and their various possible mechanisms, and also the methods of investigation employed. Finally, the behaviour of the chief nuclear materials heated in the various gases is studied successively. Materials used for fuel (metallic uranium, uranium oxide, carbides and silicides), canning materials (magnesium, aluminium, zirconium, beryllium, stainless and refractory steels), structural materials (ordinary or slightly alloyed steels), and finally moderators (graphite, beryllium oxide) are deal with in this way. This account is backed up both by the results obtained at the CEA and by work published outside or abroad up to the present day. In conclusion, every effort has been made to direct future research on the basis of the foregoing. Reprint of a paper published in Industries Atomiques - no. 9/10, 1959, p. 3-23 [fr

  2. SSTAR: The US lead-cooled fast reactor (LFR)

    International Nuclear Information System (INIS)

    Smith, Craig F.; Halsey, William G.; Brown, Neil W.; Sienicki, James J.; Moisseytsev, Anton; Wade, David C.

    2008-01-01

    It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system

  3. Lead-cooled flexible conversion ratio fast reactor

    International Nuclear Information System (INIS)

    Nikiforova, Anna; Hejzlar, Pavel; Todreas, Neil E.

    2009-01-01

    Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO 2 (S-CO 2 ) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO 2 PCS.

  4. 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.

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

    International Nuclear Information System (INIS)

    1977-01-01

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

  6. Conceptual design of a hydrogen production system by DME steam reforming and high-efficiency nuclear reactor technology

    International Nuclear Information System (INIS)

    Fukushima, Kimichika; Ogawa, Takashi

    2003-01-01

    Hydrogen is a potential alternative energy source and produced commercially by methane (natural gas) or LPG steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, since this process emits large amounts of CO 2 , replacement of the combustion heat source with a nuclear heat source for 773-1173 K processes has been proposed in order to eliminate these CO 2 emissions. This paper proposes a novel method of low-temperature nuclear hydrogen production by reforming dimethyl ether (DME) with steam produced by a low-temperature nuclear reactor at about 573 K. The authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573 K. By setting this low-temperature hydrogen production process at about 573K upstream from a turbine, it was found theoretically that the total energy utilization efficiency is about 50% and very high. By setting a turbine upstream of the hydrogen production plant, an overall efficiency of is 75% for an FBR and 76% for a supercritical-water cooled power reactor (SCPR). (author)

  7. Auxiliary equipment cooling circuit in nuclear reactors

    International Nuclear Information System (INIS)

    Yanagisawa, Ko.

    1986-01-01

    Purpose: To prevent the propagation of bacterias that transform NO 2 into NO 3 in auxiliary equipment coolants using corrosion inhibitors of nitrite type in BWR type reactors. Method: In auxiliary equipments coolant systems, water quality is controlled by using purified water as supplement water and nitrite such as Na 2 NO 2 as the corrosion inhibitors. However, in the circumstance where dissolved oxygen is present, bacteria propagate to oxidize NO 2 into NO 3 . Thus, NO 2 at 200 ppm is reduced to 20 ppm. In view of the above, a surge tank supplied from water supplement line is connected in series and a deaeration device is disposed thereto. Since the presence of dissolved oxygen causes the bacteria to propagate it is desired that the dissolved oxygen density in the supplement water is less than 5 ppm. Deaeration and pressure reduction in the surge tank can remove the dissolved oxygen, prevent NO 3 increase and also prevent stress corrosion cracks in the system pipeways. (Horiuchi, T.)

  8. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.; Stipan, L.

    1992-03-01

    A survey of steam generator operating experience for 1986 has been carried out for 184 pressurized water and pressurized heavy-water reactors, and 1 water-cooled, graphite-moderated reactor. Tubes were plugged at 75 of the reactors (40.5%). In 1986, 3737 tubes were plugged (0.14% of those in service) and 3148 tubes were repaired by sleeving. A small number of reactors accounted for the bulk of the plugged tubes, a phenomenon consistent with previous years. For 1986, the available tubesheet sludge data for 38 reactors has been compiled into tabular form, and sludge/deposit data will be incorporated into all future surveys

  9. Some aspects affecting fast reactor steam generator integrity considered from a utility viewpoint

    Energy Technology Data Exchange (ETDEWEB)

    Bolt, P R

    1975-07-01

    The important conditions affecting fast reactor steam generator integrity are discussed. In addition to the need for high integrity levels when the steam generator is first delivered to the power station site, the equally important aspect of demonstrating retention of continued high levels of integrity throughout the operating life of the station is described. The functional and related conditions that are believed important to the selection of a design type which can offer adequately high levels of integrity are given. Some of the data needs of a utility concerned with fast reactor S.G.U. design assessment are described, particular emphasis being given to areas believed to have a significant effect on steam generator reliability and integrity. (author)

  10. 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

  11. Study on the nuclear heat application system with a high temperature gas-cooled reactor and its safety evaluation (Thesis)

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo

    2008-03-01

    Aiming at the realization of the nuclear heat application system with a High Temperature Gas-cooled Reactor (HTGR), research and development on the whole evaluation of the system, the connection technology between the HTGR and a chemical plant such as the safety evaluation against the fire and explosion and the control technology, and the vessel cooling system of the HTGR were carried out. In the whole evaluation of the nuclear heat application system, an ammonia production system using nuclear heat was examined, and the technical subjects caused by the connection of the chemical plant to the HTGR were distilled. After distilling the subjects, the safety evaluation method against the fire and explosion to the reactor, the mitigation technology of thermal disturbance to the reactor, and the reactor core cooling by the vessel cooling system were discussed. These subjects are very important in terms of safety. About the fire and explosion, the safety evaluation method was established by developing the process and the numerical analysis code system. About the mitigation technology of the thermal disturbance, it was demonstrated that the steam generator, which was installed at the downstream of the chemical reactor in the chemical plant, could mitigate the thermal disturbance to the reactor. In order to enhance the safety of the reactor in accidents, the heat transfer characteristic of the passive indirect core cooling system was investigated, and the heat transfer equation considering both thermal radiation and natural convection was developed for the system design. As a result, some technical subjects related to safety in the nuclear heat application system were solved. (author)

  12. Design study on sodium cooled large-scale reactor

    International Nuclear Information System (INIS)

    Murakami, Tsutomu; Hishida, Masahiko; Kisohara, Naoyuki

    2004-07-01

    In Phase 1 of the 'Feasibility Studies on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2, design improvement for further cost reduction of establishment of the plant concept has been performed. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2003, which is the third year of Phase 2. In the JFY2003 design study, critical subjects related to safety, structural integrity and thermal hydraulics which found in the last fiscal year has been examined and the plant concept has been modified. Furthermore, fundamental specifications of main systems and components have been set and economy has been evaluated. In addition, as the interim evaluation of the candidate concept of the FBR fuel cycle is to be conducted, cost effectiveness and achievability for the development goal were evaluated and the data of the three large-scale reactor candidate concepts were prepared. As a results of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  13. Solar membrane natural gas steam-reforming process: evaluation of reactor performance

    NARCIS (Netherlands)

    de Falco, M.; Basile, A.; Gallucci, F.

    2010-01-01

    In this work, the performance of an innovative plant for efficient hydrogen production using solar energy for the process heat duty requirements has been evaluated via a detailed 2D model. The steam-reforming reactor consists of a bundle of coaxial double tubes assembled in a shell. The annular

  14. Solar membrane natural gas steam-reforming process : evaluation of reactor performance

    NARCIS (Netherlands)

    Falco, de M.; Basile, A.; Gallucci, F.

    2010-01-01

    In this work, the performance of an innovative plant for efficient hydrogen production using solar energy for the process heat duty requirements has been evaluated via a detailed 2D model. The steam-reforming reactor consists of a bundle of coaxial double tubes assembled in a shell. The annular

  15. Ethanol steam reforming kinetics of a Pd-Ag membrane reactor

    NARCIS (Netherlands)

    Tosti, S.; Basile, A.; Borelli, R.; Borgognoni, F.; Castelli, S.; Fabbricino, M.; Gallucci, F.; Licusati, C.

    2009-01-01

    The ethanol steam reforming reaction carried out in a Pd-based tubular membrane reactor has been modelled via a finite element code. The model considers the membrane tube divided into finite volume elements where the mass balances for both lumen and shell sides are carried out accordingly to the

  16. Opinion on serviceability of Bugey 3 reactor steam generators until their replacement foreseen in September 2010

    International Nuclear Information System (INIS)

    2010-04-01

    This document briefly reports the damage characterization of tubular bundles in steam generators of the Bugey 3 reactor, discusses the actions which are foreseen to prevent a tube failure risk, and discusses the risk of leakage during operation. Recommendations are formulated about investigation on the corrosion, and about prediction computation to be performed

  17. Analysis on blow-down transient in water ingress accident of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang, Yan; Zheng, Yanhua; Li, Fu; Shi, Lei

    2014-01-01

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor, which will cause a positive reactivity introduction with the increase of steam density in reactor core to enhance neutron slowing-down, also the chemical corrosion of graphite fuel elements and the damage of reflector structure material. The increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The research on water ingress transient is significant for the verification of inherent safety characteristics of high temperature gas-cooled reactor. The 200 MWe high temperature gas-cooled reactor (HTR-PM), designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is exampled to be analyzed in this paper. The design basis accident (DBA) scenarios of double-ended guillotine break of single heat-exchange tube (steam generator heat-exchange tube rupture) are simulated by the thermal-hydraulic analysis code, and some key concerns which are relative to the amount of water into the reactor core during the blow-down transient are analyzed in detail. The results show that both of water mass and steam ratio of the fluid spouting from the broken heat-exchange tube are affected by break location, which will increase obviously with the broken location closing to the outlet of the heat-exchange tube. The double-ended guillotine rupture at the outlet of the heat-exchange will result more steam penetrates into the reactor core in the design basis accident of water ingress. The mass of water ingress will also be affected by the draining system. It is concluded that, with reasonable optimization on design to balance safety and economy, the total mass of water ingress into the primary circuit of reactor could be limited effectively to meet the safety requirements, and the pollution of

  18. 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

  19. Emergency core cooling system for a fast reactor

    International Nuclear Information System (INIS)

    Johnson, H.G.; Madsen, R.N.

    1976-01-01

    The main heat transport system for a liquid-metal-cooled nuclear reactor is constructed with elevated piping and guard vessels or pipes around all components of the system below the elevation of the elevated piping so the head developed by the pumps at emergency motor speed will be unsufficient to lift the liquid-metal-coolant over the top of the guard tanks or pipes or out of the elevated piping in the event of a loss-of-coolant accident. In addition, inlet downcomers to the reactor vessel are contained within guard standpipes having a clearance volume as small as practicable. 4 claims, 2 drawing figures

  20. 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)

  1. 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

  2. Use of superheated steam to anneal the reactor pressure vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1994-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 degrees F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 degrees F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors

  3. Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

    International Nuclear Information System (INIS)

    Boardman, J.R.

    1994-10-01

    This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures

  4. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  5. 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

  6. Economic competitiveness requirements for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Hudson, C.R.; Bertel, E.; Paik, K.H.; Roh, J.H.; Tort, V.

    1999-01-01

    This paper analyses the necessary economic conditions for evolutionary water cooled reactors to be competitive. Utilising recent national cost data for fossil-fired base load plants expected to be commissioned by 2005 -2010, target costs for nuclear power plants are discussed. Factors that could contribute to the achievement of those targets by evolutionary water cooled reactors are addressed. The feed-back from experience acquired in implementing nuclear programmes is illustrated by some examples from France and the Republic of Korea. The paper discusses the impacts on nuclear power competitiveness of globalisation and deregulation of the electricity market and privatisation of the electricity sector. In addition, issues related to external cost internalisation are considered. (author)

  7. 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)

  8. 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)

  9. 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.

  10. 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

  11. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1987-01-01

    A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

  12. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  13. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

    Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

    1982-01-01

    Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

  14. 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

  15. 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

  16. Design of sodium cooled reactor systems and components for maintainability

    International Nuclear Information System (INIS)

    Carr, R.W.; Charnock, H.O.; McBride, J.P.

    1978-09-01

    Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

  17. 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

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2011-08-01

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

  20. Heat transfer performance comparison of steam and air in gas turbine cooling channels with different rib angles

    Science.gov (United States)

    Shi, Xiaojun; Gao, Jianmin; Xu, Liang; Li, Fajin

    2013-11-01

    Using steam as working fluid to replace compressed air is a promising cooling technology for internal cooling passages of blades and vanes. The local heat transfer characteristics and the thermal performance of steam flow in wide aspect ratio channels ( W/ H = 2) with different angled ribs on two opposite walls have been experimentally investigated in this paper. The averaged Nusselt number ratios and the friction factor ratios of steam and air in four ribbed channels were also measured under the same test conditions for comparison. The Reynolds number range is 6,000-70,000. The rib angles are 90°, 60°, 45°, and 30°, respectively. The rib height to hydraulic diameter ratio is 0.047. The pitch-to-rib height ratio is 10. The results show that the Nusselt number ratios of steam are 1.19-1.32 times greater than those of air over the range of Reynolds numbers studied. For wide aspect ratio channels using steam as the coolant, the 60° angled ribs has the best heat transfer performance and is recommended for cooling design.

  1. Implementation of new core cooling monitoring system for light water reactors - BCCM (Becker Core Cooling Monitor)

    International Nuclear Information System (INIS)

    Coville, Patrick; Eliasson, Bengt; Stromqvist, Erik; Ward, Olav; Fox, Georges; Ashjian, D. T.

    1998-01-01

    Core cooling monitors are key instruments to protect reactors from large accidents due to loss of coolant. Sensors presented here are based on resistance thermometry. Temperature dependent resistance is powered by relatively high and constant current. Value of this resistance depends on thermal exchange with coolant and when water is no more surrounding the sensors a large increase of temperature is immediately generated. The same instrument can be operated with low current and will measure the local temperature up to 1260 o C in case of loss of coolant accident. Sensors are manufactured with very few components and materials already qualified for long term exposure to boiling or pressurized water reactors environment. Prototypes have been evaluated in a test loop up to 160 bars and in the Barsebaeck-1 reactor. Industrial sensors are now in operation in reactor Oskarshamn 2. (author)

  2. 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

  3. 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

  4. 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

  5. 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.

  6. Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Riley, M.P.; Mohanta, L.; Miller, D.J.; Cheung, F.B. [Pennsylvania State Univ., University Park, PA (United States); Bajorek, S.M.; Tien, K.; Hoxie, C.L. [U.S. Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    2016-07-15

    A subchannel analysis of the steam cooling data obtained in the Rod Bundle Heat Transfer (RBHT) test facility has been performed in this study to capture the effect of spacer grids on heat transfer. The RBHT test facility has a 7 x 7 rod bundle with heater rods and with seven spacer grids equally spaced along the length of the rods. A method based on the concept of momentum and heat transport analogy has been developed for calculating the subchannel bulk mean temperature from the measured steam temperatures. Over the range of inlet Reynolds number, the local Nusselt number was found to exhibit a minimum value between the upstream and downstream spacer grids. The presence of a spacer grid not only affects the local Nusselt number downstream of the grid but also affects the local Nusselt number upstream of the next grid. A new correlation capturing the effect of Reynolds number on the local flow restructuring downstream as well as upstream of the spacer grids was proposed for the minimum Nusselt number. In addition, a new enhancement factor accounting for the effects of the upstream as well as downstream spacer grids was developed from the RBHT data. The new enhancement factor was found to compare well with the data from the ACHILLLES test facility.

  7. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

    The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

  8. Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

    International Nuclear Information System (INIS)

    Ishiyama, Shintaro; Muto, Yasushi; Kato, Yasuyoshi; Nishio, Satoshi; Hayashi, Takumi; Nomoto, Yasunobu

    2008-01-01

    Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO 2 (S-CO 2 ) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480degC, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO 2 turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO 2 cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m 3 and 7240 m 3 for the steam turbine system and S-CO 2 turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO 2 than in H 2 O. Therefore, the S-CO 2 turbine system is recommended to the fusion reactor system than the steam turbine system. (author)

  9. 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

  10. Testing and analyses of a high temperature thermal barrier for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.; Felten, P.

    1979-01-01

    A full size, multi-panel section of a thermal barrier system was fabricated from a nickel-base superalloy and a combination of fibrous blanket insulation materials for specific application in a steam cycle gas-cooled nuclear reactor. The 2.4 m square array was representative of the sidewall of the lower core outlet plenum and included coverplates, attachments, seals, and a simulated water-cooled liner. Testing was conducted in a reactor grade, helium-filled chamber at 816 0 C for 100 hours, which established a normal (baseline) condition; 982 0 C for 10 hours, which satisfied an emergency condition; 1093 0 C for 1 hour, which simulated a faulted condition; and 1260 0 C, which was a non-design condition test to demonstrate the temperature overshoot capability of the system. Post-test examination indicated: (1) an acceptable performance by the anti-friction chromium carbide (Cr 3 C 2 ) coating; (2) no significant galling between non-coated surfaces; (3) no distortion of attachment fixtures; (4) predictable coverplate deflection during the design conditions testing (normal, emergency, and faulted); and (5) considerable plastic deformation resulting from the near-incipient melting temperature. (orig.)

  11. 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

  12. Thermophysical properties of materials for water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA`s International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs.

  13. Thermophysical properties of materials for water cooled reactors

    International Nuclear Information System (INIS)

    1997-06-01

    The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs

  14. The gas-cooled high temperature reactor: perspectives, problems and programmes

    International Nuclear Information System (INIS)

    Beckurts, K.H.; Engelmann, P.; Erb, D.E.

    1977-01-01

    For nearly 20 years, extensive research and development programs on Helium-cooled high-temperature reactors (HTR) have been carried out in several countries of the world, in particular in Germany and in the United States. This reactor system offers major potential advantages as a source of electricity or of nuclear process heat: it shows high nuclear fuel conversion efficiency, permitting a better utilization of uranium and in particular of thorium resources; it offers a high degree of inherent nuclear safety and thus a good potential for adoption to very strict safety standards; it permits high-efficiency electricity generation using either the indirect steam or the direct Helium cycle; dry air cooling can be employed without major economic penalties; it permits direct use of the nuclear heat for the production of gaseous or liquid secondary fuels from coal and other fossil fuels or - on a more extended time scale - by thermochemical water splitting. As a result of the longstanding efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high-temperature fuels and materials technology. Three small experimental plants - Peach Bottom in USA, Dragon in England, and AVR in Germany - have been operated successfully over extended periods of time. The AVR is still in operation; since 1974 it has performed satisfactorily with an average gas outlet temperature of 950 0 C. Prototype steam-cycle plants of 300 MW(e) are underway at Fort St. Vrain, USA (full-power operation scheduled for 1977), and at Schmehausen, Germany (scheduled for 1979). Major delays have occured in the construction and commissioning of these plants; they are due to various reasons and do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successfull. Major effects by both government and industry are

  15. 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.

  16. 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

  17. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

    Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

  18. 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

  19. High power density reactors based on direct cooled particle beds

    International Nuclear Information System (INIS)

    Powell, J.R.; Horn, F.L.

    1985-01-01

    Reactors based on direct cooled HTGR type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out long the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBR's) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed. 12 figs

  20. 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)