WorldWideScience

Sample records for air cooled reactors

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

    International Nuclear Information System (INIS)

    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

  2. Temperature distribution in graphite during annealing in air cooled reactors

    International Nuclear Information System (INIS)

    A model for the evaluation temperature distributions in graphite during annealing operation in graphite. Moderated an-cooled reactors, is presented. One single channel and one dimension for air and graphite were considered. A numerical method based on finite control volumes was used for partioning the mathematical equations. The problem solution involves the use of unsteady equations of mass, momentum and energy conservation for air, and energy conservation for graphite. The source term was considered as stored energy release during annealing for describing energy conservation in the graphite. The coupling of energy conservation equations in air and graphite is performed by the heat transfer term betwen air and graphite. The results agree with experimental data. A sensitivity analysis shown that the termal conductivity of graphite and the maximum inlet channel temperature have great effect on the maximum temperature reached in graphite during the annealing. (author)

  3. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  4. Design guide for category VI reactors: air-cooled graphite reactors

    International Nuclear Information System (INIS)

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned air-cooled graphite reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission

  5. Preliminary Design of KAIST Micro Modular Reactor with Dry Air Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Baik, Seung Joon; Bae, Seong Jun; Kim, Seong Gu; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    KAIST research team recently proposed a Micro Modular Reactor (MMR) concept which integrates power conversion unit (PCU) with the reactor core in a single module. Using supercritical CO{sub 2} as a working fluid of cycle can achieve physically compact size due to small turbomachinery and heat exchangers. The objective of this project is to develop a concept that can operate at isolated area. The design focuses especially on the operation in the inland area where cooling water is insufficient. Thus, in this paper the potential for dry air cooling of the proposed reactor will be examined by sizing the cooling system with preliminary approach. The KAIST MMR is a recently proposed concept of futuristic SMR. The MMR size is being determined to be transportable with land transportation. Special attention is given to the MMR design on the dry cooling, which the cooling system does not depend on water. With appropriately designed air cooling heat exchanger, the MMR can operate autonomously. Two types of air cooling methods are suggested. One is using fan and the other is utilizing cooling tower for the air flow. With fan type air cooling method it consumes about 0.6% of generated electricity from the nuclear reactor. Cooling tower occupies an area of 227 m{sup 2} and 59.6 m in height. This design is just a preliminary estimation of the dry cooling method, and therefore more detailed and optimal design will be followed in the next phase.

  6. Preliminary Design of KAIST Micro Modular Reactor with Dry Air Cooling

    International Nuclear Information System (INIS)

    KAIST research team recently proposed a Micro Modular Reactor (MMR) concept which integrates power conversion unit (PCU) with the reactor core in a single module. Using supercritical CO2 as a working fluid of cycle can achieve physically compact size due to small turbomachinery and heat exchangers. The objective of this project is to develop a concept that can operate at isolated area. The design focuses especially on the operation in the inland area where cooling water is insufficient. Thus, in this paper the potential for dry air cooling of the proposed reactor will be examined by sizing the cooling system with preliminary approach. The KAIST MMR is a recently proposed concept of futuristic SMR. The MMR size is being determined to be transportable with land transportation. Special attention is given to the MMR design on the dry cooling, which the cooling system does not depend on water. With appropriately designed air cooling heat exchanger, the MMR can operate autonomously. Two types of air cooling methods are suggested. One is using fan and the other is utilizing cooling tower for the air flow. With fan type air cooling method it consumes about 0.6% of generated electricity from the nuclear reactor. Cooling tower occupies an area of 227 m2 and 59.6 m in height. This design is just a preliminary estimation of the dry cooling method, and therefore more detailed and optimal design will be followed in the next phase

  7. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  8. Indirect air cooling techniques for control rod drives in the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The high temperature engineering test reactor (HTTR) is the first high-temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 deg. C and thermal power of 30 MW. Sixteen pairs of control rods are employed for controlling the reactivity change of the HTTR. Each standpipe for a pair of the control rods, which is placed on the top head dome of the reactor pressure vessel, contains one control rod drive mechanism. The control rod drive mechanism may malfunction because of reduction of the electrical insulation of the electromagnetic clutch when the temperature exceeds 180 deg. C. Because 31 standpipes stand close together in the standpipe room, 16 standpipes for the control rods, which are located at the center, should be cooled effectively. Therefore, the control rod drives are cooled indirectly by forced air circulation through a pair of ring-ducts with proper air outlet nozzles and inlets. Based on analytical results, a pair of the ring-ducts was installed as one of structures in the standpipe room. Evaluation results through the rise-to-power test of the HTTR showed that temperatures of the electromagnetic clutch and the ambient helium gas inside the control rod standpipe should be below the limits of 180 and 75 deg. C, respectively, at full power operation and at the scram from the operation.

  9. STUDY ON AIR INGRESS MITIGATION METHODS IN THE VERY HIGH TEMPERATURE GAS COOLED REACTOR (VHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh

    2011-03-01

    An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident by using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (4) preventing fluid density gradient; (5) preventing fluid temperature gradient; (6) preventing high temperature. Based on the basic concepts listed above, various air

  10. Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-05-15

    A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized.

  11. Gas-cooled reactors

    International Nuclear Information System (INIS)

    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)

  12. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  13. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

    International Nuclear Information System (INIS)

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.)

  14. Air cooling system

    International Nuclear Information System (INIS)

    A procedure for cooling the steam from a turbine used in conjunction with a power nuclear reactor has been described in the main patent. According to said procedure, use is made of a circuit where a two-phase mixture is circulated, said closed circuit connecting the turbine condenser to a cooling tower. According to the present addition patent, the cooling structure is provided with cooling fins previously hollowed in view of increasing the interface between the fluid and said structure, which improves the performance of the system

  15. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    International Nuclear Information System (INIS)

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  16. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  17. High temperature gas-cooled reactor technology

    International Nuclear Information System (INIS)

    The high temperature gas-cooled reactor (HTGR) with a direct cycle helium system has drawn attention as the next generation nuclear power plant that is closest to commercialization. Fuji Electric participated in the design, manufacture and construction of JAPCO's Tokai-1 plant, a 'Colder Hall' type reactor, which was the first commercial nuclear power plant in Japan, and JAERI's high temperature engineering test reactor (HTTR), which was the first high temperature gas-cooled reactor in Japan. Fuji Electric, a pioneer of gas-cooled reactors, worked on the design, construction and development of these reactors. This paper provides brief descriptions of the air-cooled spent fuel storage system of the HTTR, material test facilities for the HTTR, and the development of an inherently safe and highly efficient commercial HTGR power plant as examples of Fuji Electric's recent activities in the HTGR field. (author)

  18. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    In the reactor operating with supercritical pressure and temperature part of the water flowing through the moderator tubes is deflected at the outlet and mixed with a residual partial flow of the coolant fed into the core as well as passed along the fuel rods in opposite direction. By special guiding of the flow downward through the guide tubes of the control rods insertion of the control rods is simplified because of reduced frictional forces. By this means it is also achieved to design less critical the control rod cooling with respect to flow rate control and operating behavior in case of a scram. (orig.)

  19. Sodium-cooled nuclear reactors

    International Nuclear Information System (INIS)

    This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index

  20. Studies of heat transfer having relevance to nuclear reactor containment cooling by buoyancy-driven air flow

    International Nuclear Information System (INIS)

    Two separate effects experiments concerned with buoyancy-influenced convective heat transfer in vertical passages which have relevance to the problem of nuclear reactor containment cooling by means of buoyancy-driven airflow are described. A feature of each is that local values of heat transfer coefficient are determined on surfaces maintained at uniform temperature. Experimental results are presented which highlight the need for buoyancy-induced impairment of turbulent convective heat transfer to be accounted for in the design of such passive cooling systems. A strategy is presented for predicting the heat removal by combined convective and radiative heat transfer from a full scale nuclear reactor containment shell using such experimental results

  1. Liquid cooled nuclear reactors

    International Nuclear Information System (INIS)

    A construction is described for a liquid metal cooled fast reactor, in which the core is supported in a pool of liquid coolant, wherein a catchment tray is provided for any debris falling from the core. The tray comprises a complex of open top collecting vessels with central support struts, the vessels being spaced apart and arranged in layers in a lattice pitch. The lattice pitches of the vessels in each layer are off-set to the lattice pitches of the vessels in the other layers, so that upper vessels partially overlap lower vessels, and the support struts extend through interspaces defined by the vessels in off-set pitch to a common supporting sub-structure. The complex of vessels offers a complete catchment area for falling debris, whilst being pervious to liquid coolant circulating upwardly by convection. The collecting vessels preferably comprise conical dishes and are arranged in triangular lattice pitch in each layer, and the complex of vessels comprises three layers. Alternatively the collecting vessels may be rectilinear and arranged on a square lattice. The catchment tray may comprise two or more such complexes in stacked array. (U.K.)

  2. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    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

  3. Cooling system for reactor container

    International Nuclear Information System (INIS)

    Purpose: To effectively cool a reactor container upon reactor shutdown with no intrusion of metal corrosion products in coolants into the main steam pipe in a BWR type reactor. Constitution: A clean up system comprising a pipeway, a recycling pump, a non-regenerative heat exchanger and a primary coolant purifier and a regenerative heat exchanger is provided branched from a residual heat removing system and the clean up system is connected by way of a valve to a feedwater pipeway, as well as connected by way of the pipeway to the main steam pipeway at the midway of two main steam separation valves outside of the reactor container. This enables to prevent metal corrosion products floating on the surface of reactor water from introducing into the main steam pipe when the pressure vessel is filled with water. Then, since the pressure vessel is filled with primary coolants, the pressure vessel can be cooled uniformly in a short time. (Ikeda, J.)

  4. Emergency cooling apparatus for reactor

    International Nuclear Information System (INIS)

    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)

  5. Cooling device upon reactor isolation

    International Nuclear Information System (INIS)

    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.)

  6. Water Cooled FBNR Nuclear Reactor

    International Nuclear Information System (INIS)

    A new era of nuclear energy is emerging through innovative nuclear reactors that are to satisfy the new philosophies and criteria that are developed by the INPRO program of the International Atomic Energy Agency (IAEA). The IAEA is establishing a new paradigm in relation to nuclear energy. The future reactors should meet the new standards in respect to safety, economy, non-proliferation, nuclear waste, and environmental impact. The Fixed Bed Nuclear Reactor (FBNR) is a small (70 MWe) nuclear reactor that meets all the established requirements. It is an inherently safe and passively cooled reactor that is fool proof against nuclear proliferation. It is simple in design and economic. It can serve as a dual purpose plant to produce simultaneously both electricity and desalinated water thus making it especially suitable to the needs of most of developing countries. FBNR is developed with the support of the IAEA under its program of Small Reactors Without On-Site Refuelling (SRWOSR). The FBNR reactor uses the pressurized water reactor (PWR) technology. It fulfills the objectives of design simplicity, inherent and passive safety, economy, standardization, shop fabrication, easy transportability and high availability. The inherent safety characteristic of the reactor dispenses with the need for containment; however, a simple underground containment is envisaged for the reactor in order to reduce any adverse visual impact. (author)

  7. Pressure drop and heat transfer in the sodium to air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    A numerical study was performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX were modeled as porous media and simulated heat and momentum transfer. Two-dimensional flow characteristic appeared at the most region of AHX annulus. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX were evaluated and compared with Zhukauskas empirical correlations. (author)

  8. Passive cooling systems in power reactors

    International Nuclear Information System (INIS)

    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

  9. Dynamic simulation of the air-cooled decay heat removal system of the German KNK-II experimental breeder reactor

    International Nuclear Information System (INIS)

    A Dump Heat Exchanger and associated feedback control system models for decay heat removal in the German KNK-II experimental fast breeder reactor are presented. The purpose of the controller is to minimize temperature variations in the circuits and, hence, to prevent thermal shocks in the structures. The basic models for the DHX include the sodium-air thermodynamics and hydraulics, as well as a control system. Valve control models for the primary and intermediate sodium flow regulation during post shutdown conditions are also presented. These models have been interfaced with the SSC-L code. Typical results of sample transients are discussed

  10. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    The gas temperature of a hot gas loop in gas-cooled nuclear reactor plants shall be able to be modified without influencing the gas temperature of the other loops. If necessary, it should be possible to stop the loop. This is possible by means of a mixer which is places below the heat absorbing component in the hot channel and which is connected to a cold gas line. (orig.)

  11. Reactor pressure vessel cooling device

    International Nuclear Information System (INIS)

    A spray pipeline for the head of a pressure vessel is connected to a spray for the pressure vessel head of the reactor pressure vessel. The pipeline is equipped with a spray flow rate control valve for controlling the flow rate of spray jetted from the head spray to the inside of the reactor and a thermometer for spray water. A reactor pressure vessel cooling and controlling portion intakes temperature signals sent from a thermometer for the head of the pressure vessel, a thermometer for the flange of the pressure vessel and a thermometer for the body of the pressure vessel, and a thermometer for spray water. Then, it outputs restriction signal for the opening degree of the spray flow rate control valve to restrict the spray water flow rate within an appropriate range when the difference between the temperature of each portion of the pressure vessel and the temperature of spray water is increased to greater than an allowable temperature difference. With such procedures, spray operation for the head of the pressure vessel can be conducted not only upon emergency but also during reactor cooling operation. (I.N.)

  12. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    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 oC 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

  13. Small power sodium cooled fast nuclear reactors

    International Nuclear Information System (INIS)

    1.5 MW(e), 12 MW(e) and 170 MW(e) small power sodium cooled fast reactors have been developed. The reactor plants were developed as universal power units for economically effective energy and industrial steam generation and heat supply. The main features increasing the power unit economic efficiency are: serial fabrication of standard RPs at the factory and delivery of reactor vessels in ready made form; realization of self-protection principles and use of passive systems in RP; use of standard machine room equipment, fabricated in accordance with the rules of conventional heat power engineering; use of turbine plant with thermodynamic coefficient, exceeding the corresponding value for the plants of PWR type. For MBRU-1.5 and MBRU-12 RPs it is proposed to use a core without FA replacement during the whole service life (30 years) and for BMN-170 RP it is proposed to use a core with a 4 year operating period and 1 year between the refueling shutdowns. During the whole service life a minimal number of operating personnel will be needed for the plant servicing. The personnel functions will be periodically to observe the parameters of technological process. Passive principles are used in the main RP safety systems: a passive type system of emergency residual heat removal system provides heat removal directly through the reactor vessel forced air cooling due to the natural air chimney effect; an emergency reactor shut-down system is provided by emergency protection rods with active-passive action. (author)

  14. Water cooled FBNR nuclear reactor

    International Nuclear Information System (INIS)

    elements from the fuel chamber up into the core. A fixed suspended core is formed in the reactor. In the shut down condition, the suspended core breaks down and the fuel elements leave the core and fall back into the fuel chamber by the force of gravity. The fuel elements are made of UO2 micro spheres embedded in zirconium and cladded by zircaloy. Any signal from any of the detectors, due to any initiating event, will cut-off power to the pump, causing the fuel elements to leave the core and fall back into the fuel chamber, where they remain in a highly subcritical and passively cooled conditions. The fuel chamber is cooled by natural convection transferring heat to the water in the tank housing the fuel chamber. The nest step in the development of FBNR is the construction of its prototype. Efforts are being made to secure participants in such an endeavor. (author)

  15. Physics of gas-cooled reactors

    International Nuclear Information System (INIS)

    This paper deals with technical aspects of the gas-cooled reactors safety and design. It content five main parts: the new safety requirements for innovative reactors; the concept of a non-meltable nuclear reactor; the new concepts of decay heat removal from the reactor system; the behaviour of reactor system in case of extreme reactivity accidents; technical innovations in the reactor technology and necessary developments and proof tests. (A.L.B.)

  16. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Reference is made to liquid metal cooled nuclear reactors of the 'pool' type. In such reactors the core, the heat exchangers, and the coolant circulating pumps are submerged in a pool of liquid metal. In operation of the reactor it is necessary to be able to locate and identify components submerged in the pool, and before moving rotating shields in the roof of the pool-containing vault it is necessary to ensure that all the normally suspended absorber rods have been inserted in the core and released from their suspensions. Television cameras are unsuitable for use in the opaque liquid metal but ultrasound in the megahertz range has been used to give a television screen kind of display. There is some difficulty, however, in transmitting ultrasound signals from a transducer into the pool of coolant because the transducer must be protected from the high temperature environment of the coolant. This difficulty has been partially overcome, however, by transmitting the signals by way of a wave guide extending from the transducer into the coolant pool. Such a wave guide may comprise a column of liquid metal within a dip tube. The column of liquid coolant is uninterrupted by a supporting diaphragm. Such a system is here described. (U.K.)

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

    International Nuclear Information System (INIS)

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

  18. Helium-cooled high temperature reactors

    International Nuclear Information System (INIS)

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg

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

    DEFF Research Database (Denmark)

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

    1976-01-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data...... extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between...... Hastelloy X and a number of other structural alloys are given....

  20. Response of fuel, fuel elements and gas cooled reactor cores under accidental air or water ingress conditions. Proceedings of a technical committee meeting held in Beijing, China, 25-27 October 1993

    International Nuclear Information System (INIS)

    The meeting was convened by the IAEA on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Netherlands, Switzerland, the Russian Federation, the United Kingdom and the United States of America. The meeting covered the following topics: experimental investigations of the effects of air and water ingress; predicted response of fuel, graphite and other reactor components; options for minimizing or mitigating the effects of air or water ingress. 19 papers were presented at the meeting. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  1. Analysis of water cooled reactors stability

    International Nuclear Information System (INIS)

    A model for stability analysis of non-boiling water cooled nuclear system is developed. The model is based on linear reactor kinetics and space averaged heat transfer in reactor and heat-exchanger. The transfer functions are defined and the analysis was applied to nuclear reactor RA at 'Boris Kidric' Institute - Vinca. (author)

  2. Advanced gas-cooled reactors (AGR)

    International Nuclear Information System (INIS)

    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

  3. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    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)

  4. Ultrasonic methodology measurement of two-phase (air-water) flows in cooling systems of nuclear reactor

    International Nuclear Information System (INIS)

    An ultrasonic methodology is proposed for the measurement for two phase (air-water) flow parameters. Ultrasonic backscattered signals were used to analyze the following parameters: average number of bubbles, interfacial area and void fraction. The results show a strong correlation between the parameters and the ultrasonic power signal obtained. (author)

  5. Air cooled absorption chillers for solar cooling applications

    Science.gov (United States)

    Biermann, W. J.; Reimann, R. C.

    1982-03-01

    The chemical composition of a 'best' absorption refrigerant system is identified, and those properties of the system necessary to design hot water operated, air cooled chilling equipment are determined. Air cooled chillers from single family residential sizes into the commercial rooftop size range are designed and operated.

  6. Gas-cooled reactors and their applications

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review and discuss the current status and recent progress made in the technology and design of gas-cooled reactors and their application for electricity generation, process steam and process heat production. The meeting was attended by more than 200 participants from 25 countries and International Organizations presenting 34 papers. The technical part of the meeting was subdivided into 7 sessions: A. Overview of the Status of Gas-Cooled Reactors and Their Prospects (2 papers); B. Experience with Gas-Cooled Reactors (5 papers); C. Description of Current GCR Plant Designs (10 papers); D. Safety Aspects (4 papers); E. Gas-Cooled Reactor Applications (3 papers); F. Gas-Cooled Reactor Technology (6 papers); G. User's Perspectives on Gas-Cooled Reactors (4 papers). At the end of the meeting a round table discussion was organized in order to summarize the meeting and to make recommendations for future activities. A separate abstract was prepared for each of the 34 presentations of this meeting. Refs, figs and tabs

  7. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    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

  8. Gas turbine modeling for NPP with helium cooled reactor

    International Nuclear Information System (INIS)

    The performance analyzes of closed helium cycle for NPPs with high-temperature gas-cooled reactors was carried out. Air-turbine units and helium-turbine units were compared. Helium turbine features were particularized in comparison with conventional air turbines. Simulation results of gas turbine setting with helium as a working medium were presented. Problems concerning high economic efficiency advance of helium turbines were discussed

  9. French activities on gas cooled reactors

    International Nuclear Information System (INIS)

    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

  10. Technological innovations for FBR reactor cooling system

    International Nuclear Information System (INIS)

    The fast breeder reactor (FBR) is expected to be commercialized early in the 21st century. In order to realize this goal, technological innovations are desired in order to extensively enhance economic performance, and improvement of the reactor cooling system is of primary importance in this regard. Over the past 10 years, Toshiba has developed a succession of new technologies in the field of reactor cooling systems, including a compact type intermediate heat exchanger (IHX), an integral once-through type steam generator (SG), a double-wall-tube type steam generator, and a sodium-immersed high-temperature type electromagnetic pump (EMP). As a synthesis of the fruits of such research and development we have formulated innovative concepts for a reactor cooling system and its constituent components. These advances in research and development activities will significantly contribute to the commercialization of FBRs. (author)

  11. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

    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. Nuclear Reactor RA Safety Report, Vol. 5, Reactor cooling systems

    International Nuclear Information System (INIS)

    RA reactor cooling system enable cooling during normal operation and under possible accidental conditions and include: technical water system, heavy water system, helium gas system, system for heavy water purification and emergency cooling system. Primary cooling system is a closed heavy water circulation system. Heavy water system is designed to enable permanent circulation and twofold function of heavy water. In the upward direction of cooling it has a coolant role and in the downward direction it is the moderator. Separate part of the primary coolant loop is the system for heavy water purification. This system uses distillation and ion exchange processes

  13. Thermal management systems for cooling nuclear reactor during emergency

    International Nuclear Information System (INIS)

    This paper discusses two different thermal management systems for nuclear reactor during emergency situation. First system will provide safer and reliable heat pipe based emergency core cooling system (ECCS) for nuclear-reactor, with initial 10 sec gravity feed water for accelerated cooling response. The designed loop type heat pipe ECCS is composed of cylindrical evaporator with 62 vertical tubes, each 150 mm diameter and 6 m length, mounted around the circumference of nuclear fuel assembly and 21 m x 10 m x 5 m naturally cooled finned condenser installed outside the primary containment. Proposed ECCS will be able to cool down core after reactor shutdown from 282degC to below 250degC within 4.3 hours of shutdown thereby providing safer environment to nuclear power plants. Second system proposes debris cooling system for nuclear reactor chamber, based on open air Bryton cycle. Such a system will provide cleaner and safer system for the nuclear reactor chamber after accident. (author)

  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. Film cooling air pocket in a closed loop cooled airfoil

    Science.gov (United States)

    Yu, Yufeng Phillip; Itzel, Gary Michael; Osgood, Sarah Jane; Bagepalli, Radhakrishna; Webbon, Waylon Willard; Burdgick, Steven Sebastian

    2002-01-01

    Turbine stator vane segments have radially inner and outer walls with vanes extending between them. The inner and outer walls are compartmentalized and have impingement plates. Steam flowing into the outer wall plenum passes through the impingement plate for impingement cooling of the outer wall upper surface. The spent impingement steam flows into cavities of the vane having inserts for impingement cooling the walls of the vane. The steam passes into the inner wall and through the impingement plate for impingement cooling of the inner wall surface and for return through return cavities having inserts for impingement cooling of the vane surfaces. To provide for air film cooing of select portions of the airfoil outer surface, at least one air pocket is defined on a wall of at least one of the cavities. Each air pocket is substantially closed with respect to the cooling medium in the cavity and cooling air pumped to the air pocket flows through outlet apertures in the wall of the airfoil to cool the same.

  16. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    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

  17. Design challenges for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    It is of vital importance for commercialized fast reactor to achieve component design with excellent integrity and economics. In the phase II of feasibility study till 2005, a system design for commercialized fast reactor for sodium cooling was achieved. For economical improvement, the system design was undertaken along the guideline including innovative technology for system simplification and new material development. In this paper, the results from the design for shortening of cooling pipings, new components and three dimensional seismic isolation are described, which are design challenges for the sodium cooled fast reactor. Furthermore, in-service inspection and repair is mentioned. Finally, economics for the simplification and the mass reduction employing above technologies are examined

  18. Liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hydrogen can be added to nuclear reactors with a liquid hydrogen-containing coolant on the suction side of a high pressure pump in the purification system. According to the invention this is performed by means of a liquid jet condenser which uses the coolant as liquid and which is preferably charged from the pressure side of the high pressure pump and conveys the liquid to a mixer connected in series with the high pressure pump. The invention is to be used especially in pressurized water reactors. (orig./PW)

  19. Liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hydrogen can be added to nuclear reactors with a liquid hydrogen-containing coolant on the suction side of a high pressure pump in the purification system. According to the invention this is performed by means of a liquid jet condenser which uses the coolant as liquid and which is preferably charged from the pressure side of the high pressure pump and conveys the liquid to a mixer connected in series with the high pressure pump. The invention is to be used especially in pressurized water reactors. (orig.)

  20. Passive cooling in modern nuclear reactors

    International Nuclear Information System (INIS)

    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

  1. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    A liquid metal cooled nuclear reactor construction is described comprising a reactor core submerged in a pool of liquid metal coolant contained in a vessel which is housed in a concrete containment vault, the roof structure of the vault having thermal insulation comprising a series of super-imposed spaced plates, with baffles disposed so as to restrict convectional flow of metal vapour through the interspaces of the plates and between the uppermost plate or plates and the vault roof structure. (author)

  2. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    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.

  5. Mixed oxide fuel for water cooled reactors

    International Nuclear Information System (INIS)

    The problems connected with introduction of plutonium extracted from spent fuels of operating NPPs into water cooled reactor fuel cycle are considered. The trends in formation of the World market of mixed fuel are illustrated taking as examples Great Britain and Japan

  6. Coolant technology of water cooled reactors. V. 1: Chemistry of primary coolant in water cooled reactors

    International Nuclear Information System (INIS)

    This report is a summary of the work performed within the framework of the Coordinated Research Programme on Investigations on Water Chemistry Control and Coolant Interaction with Fuel and Primary Circuit Materials in Water Cooled Power Reactors organized by the IAEA and carried out from 1987 to 1991. It is the continuation of a programme entitled Reactor Water Chemistry Relevant to Coolant-Cladding Interaction (IAEA-TECDOC-429), which ran from 1981 to 1986. Subsequent meetings resulted in the title of the programme being changed to Coolant Technology of Water Cooled Reactors. The results of this Coordinated Research Programme are published in four volumes with an overview in the Technical Reports Series. The titles of the volumes are: Volume 1: Chemistry of Primary Coolant in Water Cooled Reactors; Volume 2: Corrosion in the Primary Coolant Systems of Water Cooled Reactors; Volume 3: Activity Transport Mechanisms in Water Cooled Reactors; Volume 4: Decontamination of Water Cooled Reactors. These publications should be of interest to experts in water chemistry at nuclear power plants, experts in engineering, fuel designers, research and development institutes active in the field and to consultants to these organizations. Refs, figs and tabs

  7. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  9. Emergency reactor core cooling system of BWR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an emergency reactor core cooling system which can reduce a capacity of a power source required upon occurrence of emergency, extending an start-up time of an emergency reactor core cooling system (ECCA) to provide a plant endurable to a common factor accident and can provide time margin up to the start-up time. Namely, the system of the present invention comprises a division I equipped with an isolation condenser (IC), an after-heat removing system (low pressure system)(LPFL/RHR) and an emergency gas turbine generator (GT), a division II equipped with a diesel driving water injection system (high pressure system)(HDIS), LPFL/RHR, and GT, and a division III equipped with a reactor isolation time cooling system (high pressure system)(ARCIC), LPFL/RHR and GT. With such a constitution, since the IC, HDIS and ARCIC are used in combination as a high pressure system, an electromotive pump required to be operated upon high pressure state can be saved. In addition, if a static reactor cooling system (PCCS) is adopted and is provided with a back-up function for LPFL/RHR with respect to heat removal of the container upon occurrence of an accident, the countermeasure for occurrence of severe accidents can be enhanced. (I.S.)

  10. Extended Cooling System for High Power Reactors

    International Nuclear Information System (INIS)

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants (NPPs) and proposed for advanced light water reactors (LWRs). However, it is not clear that currently proposed external reactor vessel cooling (ERVC) could provide sufficient heat removal for higher power reactors. This paper proposes a dual retention strategy to realize fail-proof defense-in-depth in the APR1400 (Advanced Power Reactor 1400 MWe) and the OPR 1000 (Optimized Power Reactor 1000 MWe). The dual retention has the advantage of IVR-ERVC as well as ex-vessel cooling (EVC) strategies. The multilateral, multidisciplinary project calls for national and international cutting-edge technologies to research and produce (R and P) the D2R2 (Duel Retention Demonstration Reactor) equipped with OASIS (Optimized Advanced Safety Injection System) and ROSIS (Reactor Outer Safety Injection System) to cope with design-basis accidents and beyond in a coherent, continual, comprehensive manner. The enterprise aims to develop the design-basis and severe accident engineering solutions. The enterprise aims to develop the design-basis and severe accident engineering solutions. The former embraces ISAIAH (Injection System Annular Interactive Aero Hydrodynamics) and MESIAH (Methodical Evaluation System Interactive Aero Hydrodynamics). The latter comprises GODIVA (Geo metrics of Direct Injection Versatile Arrangement), SONATA (Simulation of Narrow Annular Thermomechanical Arrest or), TOCATA (Termination of Corium Ablation Thermal Attack) and STRADA (Solution to Reactor Advanced Design Alternatives). D2R2 will contribute to enhancement of both safety and economics for an advanced high power particular and nuclear power in general

  11. Startup of the FFTF sodium cooled reactor

    International Nuclear Information System (INIS)

    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

  12. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    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)

  13. CEA programme on gas cooled reactors

    International Nuclear Information System (INIS)

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

  14. Emergency cooling system for nuclear reactors

    International Nuclear Information System (INIS)

    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

  15. The development of air cooled condensation systems

    International Nuclear Information System (INIS)

    EGI - Contracting/Engineering has had experience with the development of air cooled condensing systems since the 1950's. There are two accepted types of dry cooling systems,the direct and the indirect ones. Due to the fact that the indirect system has several advantages over the direct one, EGI's purpose was to develop an economic, reliable and efficient type of indirect cooling system, both for industrial and power station applications. Apart from system development, the main components of dry cooling plant have been developed as well. These are: the water-to-air heat exchangers; the direct contact (DC, or jet) condenser; the cooling water circulating pumps and recovery turbines; and the peak cooling/preheating units. As a result of this broad development work which was connected with intensive market activity, EGI has supplied about 50% of the dry cooling plants employed for large power stations all over the world. This means that today the cumulated capacity of power units using Heller type dry cooling systems supplied and contracted by EGI is over 6000 MW

  16. Analysis of two phase natural circulation flow in the reactor cavity under external reactor vessel cooling

    International Nuclear Information System (INIS)

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR (Advanced Power Reactor) 1400, a K-HERMES-HALF (Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale) experiment was performed at KAERI (Korea Atomic Energy Research Institute) using an air injection method. This experiment was analyzed to verify and evaluate the experimental results using the RELAP5/MOD3 computer code. In addition, the geometry scaling on full height & full sector, and a material scaling between air-water and steam-water two phase natural circulation flow, have been performed for an application of the experimental results to an actual APR1400. The RELAP5/MOD3 results on the water circulation mass flow rate are very similar to the experimental results, in general. The water circulation mass flow rate of the full height & full sector case is approximately 7.6-times higher than that of the K-HERMEL-HALF case. The water circulation mass flow rate of the air injection case is 20-50 % higher than that of the steam injection case at 20 % of the injection rate. (author)

  17. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    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

  18. Bacterial pathogens in a reactor cooling reservoir

    International Nuclear Information System (INIS)

    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

  19. Design codes for gas cooled reactor components

    International Nuclear Information System (INIS)

    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

  20. Design study of cooling system for tokamak fusion reactor

    International Nuclear Information System (INIS)

    Design study of the reactor cooling system for a tokamak fusion reactor has been carried out. In the cooling system of an experimental 150 MWt fusion reactor, to grasp the plant concept and clarify the R and D items the main cooling system and the tritium recovery system were designed and the auxiliary system was examined. In the cooling system of a commercial 2000 MWt fusion reactor, to study the plant and environment safety the main cooling system and the tritium recovery system were designed, including the evaluation of water leakage and tritium penetration in the steam generators. (auth.)

  1. A Blast of Cool Air

    Science.gov (United States)

    2000-01-01

    Unable to solve their engineering problem with a rotor in their Orbital Vane product, DynEco Corporation turned to Kennedy Space Center for help. KSC engineers determined that the compressor rotor was causing a large concentration of stress, which led to cracking and instant rotor failure. NASA redesigned the lubrication system, which allowed the company to move forward with its compressor that has no rubbing parts. The Orbital Vane is a refrigerant compressor suitable for mobile air conditioning and refrigeration.

  2. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

    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)

  3. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  4. Lessons learned from Gen-I carbon dioxide cooled reactors

    International Nuclear Information System (INIS)

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21. century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors. (author)

  5. Emergency cooling of pressurized water reactors

    International Nuclear Information System (INIS)

    The operating conditions of the reactor are first described during emergency cooling (system aspect) taking as example a Westinghouse type reactor (cold branch injection). Mention is then made of the different variants employed either by Westinghouse or by other manufacturers (top of vessel injection, hot branch injection, equalizing valve, etc.). The importance of accurately knowing the pressure drops throughout the system is demonstrated. Particular attention is given to heat exchanges in the core: non-wet area, heat transfer by convection and radiation with a two-phase flow; re-wetting area, effect of axial conduction and of the condition of the fuel. Finally, the models used for the overall study of the system and for the treatment of local heat transfer, are briefly presented

  6. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    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)

  7. Design of a natural draft air-cooled condenser and its heat transfer characteristics in the passive residual heat removal system for 10 MW molten salt reactor experiment

    International Nuclear Information System (INIS)

    As one of the Generation IV reactors, Molten Salt Reactor (MSR) has its superiorities in satisfying the requirements on safety. In order to improve its inherent safety, a concept of passive residual heat removal system (PRHRS) for the 10 MW Molten Salt Reactor Experiment (MSRE) was put forward, which mainly consisted of a fuel drain tank, a feed water tank and a natural draft air-cooled condenser (NDACC). Besides, several valves and pipes are also included in the PRHRS. A NDACC for the PRHRS was preliminarily designed in this paper, which contained a finned tube bundle and a chimney. The tube bundle was installed at the bottom of the chimney for increasing the velocity of the air across the bundle. The heat transfer characteristics of the NDACC were investigated by developing a model of the PRHRS using C++ code. The effects of the environmental temperature, finned tube number and chimney height on heat removal capacity of the NDACC were analyzed. The results show that it has sufficient heat removal capacity to meet the requirements of the residual heat removal for MSRE. The effects of these three factors are obvious. With the decay heat reducing, the heat dissipation power declines after a short-time rise in the beginning. The operation of the NDACC is completely automatic without the need of any external power, resulting in a high safety and reliability of the reactor, especially once the accident of power lost occurs to the power plant. - Highlights: • A model to study the heat transfer characteristics of the NDACC was developed. • The NDACC had sufficient heat removal capacity to remove the decay heat of MSRE. • NDACC heat dissipation power depends on outside temperature and condenser geometry. • As time grown, the effects of outside temperature and condenser geometry diminish. • The NDACC could automatically adjust its heat removal capacity

  8. Thermoelectric Air Cooling For Cars

    Directory of Open Access Journals (Sweden)

    Manoj S. Raut

    2012-05-01

    Full Text Available India is the second most populous country in the world with over 1.21billion people (estimated for April, 2011,more than sixth of the world’s population. India is projected to be world’s most populous country by 2025,surpassing china, its population exceeding 1.6 billion people by 2050.Comparing with the population there are 2.65 million cars sold in India as of march 2011.According to the society of Indian automotive manufacturer, annual car sales are projected to increase up to 5 million vehicles by 2015 and more than a 9 million by 2020.By 2050,the country is expected to top of the world in car volumes with approximately 611 million vehicles on the nation’s roads.The above data shows that, as the population increase the no. of vehicles also increase. Today, an automobile is a necessity for everyone. For a long or short journey people need car regard to thesafety, environment and most important comfort. Owing to these reasons, many vehicles are equipped with heating, ventilating and air conditioning system. In today’s world, no one feel comfortable in a vehicle without HVAC system. Therefore, HVAC becomes an integral part of human life. Today’s present HVAC system is very efficient and reliable but it has some demerits. It has been observed during the last two decades that the O3 –layer is slowly destroyed because of the refrigerant (CFC and HFC used for the refrigeration and air –conditioning purposes. The common refrigerant used is HFC’s which are leaked and slowly climb into the atmosphere. When they reach to O3 layer they act on O3 –molecules and the layer of O3 is destroyed.

  9. LFR "Lead-Cooled Fast Reactor"

    Energy Technology Data Exchange (ETDEWEB)

    Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C; Suh, K

    2006-05-11

    The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small, transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology

  10. 04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

    International Nuclear Information System (INIS)

    The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

  11. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

    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

  12. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    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.

  13. Status of and prospects for gas-cooled reactors

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-10-01

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

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

    International Nuclear Information System (INIS)

    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)

  17. Status of national gas cooled reactor programmes

    International Nuclear Information System (INIS)

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

  18. Technological readiness of evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    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)

  19. Anomalous Effects in Air While Cooling Water

    CERN Document Server

    Sardo, Rachel

    2008-01-01

    Water is a unique compound with many anomalies and properties not fully understood. Designing an experiment in the laboratory to study such anomalies, we set up a series of experiments where a tube was placed inside a sealed container with thermocouples attached to the outer surface of the tube and in the air adjacent to the tube. Alternately, deionized water and other compounds were added to the tube and cooled to freezing. Several of the thermocouples suspended in the air and adjacent to the tube showed thermal oscillations as the overall temperature of the container was decreasing. The temperature of the thermocouples increased and decreased in a sinusoidal way during part of the cool down to freezing. Thermal oscillations as large as 3 degrees Celsius were recorded with typical frequencies of about 5 oscillations per minute.

  20. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    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. 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)

  2. Comparison of Air Cooled and Evaporatively Cooled Refrigerartion Systems – A Review Paper

    OpenAIRE

    V. V. Birangane; A.M.Patil

    2014-01-01

    The air cooled condensers are widely used as they are less costly and give satisfactory performance. But their performance is greatly affected by the temperature of cooling media which is ambient air. To deal this problem we can use water cooled condenser. But their cost and maintenance limit their use. The performance improvement of Air cooled condensers can be achieved by using evaporative cooling. This method may prove quiet effective and less costly. There are researchers ...

  3. An Innovative Reactor Technology to Improve Indoor Air Quality

    Energy Technology Data Exchange (ETDEWEB)

    Rempel, Jane [TIAX LLC., Lexington, MA (United States)

    2013-03-30

    As residential buildings achieve tighter envelopes in order to minimize energy used for space heating and cooling, accumulation of indoor air pollutants such as volatile organic compounds (VOCs), becomes a major concern causing poor air quality and increased health risks. Current VOC removal methods include sorbents, ultraviolet photocatalytic oxidation (UVPCO), and increased ventilation, but these methods do not capture or destroy all VOCs or are prohibitively expensive to implement. TIAX's objective in this program was to develop a new VOC removal technology for residential buildings. This novel air purification technology is based on an innovative reactor and light source design along with UVPCO properties of the chosen catalyst to purify indoor air and enhance indoor air quality (IAQ). During the program we designed, fabricated and tested a prototype air purifier to demonstrate its feasibility and effectiveness. We also measured kinetics of VOC destruction on photocatalysts, providing deep insight into reactor design.

  4. On the Method of Air Jet Cooling in Green Manufacturing

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Green cooling is an important technology in green manufacturing. In the way of jetting, cooling airflow is used in the experiments of metal material cutting, by compari- son of the changes of some technological factors, such as cutting heat, surface finish, in the process of jet cooling, pour cooling and natural cooling, we can draw the conclusion that air jet cooling has a better cooling effect and green function. It can be widely used in both aditional and automatic green manufacturing.

  5. Nanostructured Catalytic Reactors for Air Purification Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR Phase II project proposes the development of lightweight compact nanostructured catalytic reactors for air purification from toxic gaseous organic...

  6. Nanostructured Catalytic Reactors for Air Purification Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR Phase I project proposes the development of lightweight compact nanostructured catalytic reactors for air purification from toxic gaseous organic...

  7. Pre-Analysis of Triga Mark II Reactor Cooling System

    OpenAIRE

    AKAY, Orhan Erdal

    2012-01-01

    In this study, work of the reactor cooling system is divided into two time zone. The second cooling circuit has been that the conditions required operating. Cooling system which is the center of the heat exchanger total heat transfer coefficient correlations were calculated using the theoretical. The design values were compared with results obtained by calculation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  9. Development of studies on helium cooled fast reactors

    International Nuclear Information System (INIS)

    A necessity is shown of developing breeders with high reproductive properties. Helium cooled fast reactor is considered. The reactor performances, heating circuit with the use of a steam turbine unit in the secondary circuit is outlined. The reactor design and fuel assemblies are described

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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. Their effective heat transport combines with the large heat-transfer coefficients of tube banks, thereby reducing 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)

  12. Cooling load differences between radiant and air systems

    OpenAIRE

    Feng, Jingjuan Dove; Schiavon, Stefano; Bauman, Fred

    2013-01-01

    Unlike the case of air systems where the cooling load is purely convective, the cooling load for radiant systems consists of both convective and radiant components. The main objectives of this energy simulation study were to investigate whether the same design cooling load calculation methods can be used for radiant and air systems by studying the magnitude of the cooling load differences between radiant and air systems over a range of configurations and to suggest potential improvem...

  13. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, 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.

  14. Analysis of Air Cooling Performance in the Coil Stack Assembly for APR+ CEDMs

    International Nuclear Information System (INIS)

    The control element drive mechanisms (CEDMs) are electromechanical devices for withdrawing, inserting, holding or tripping the control element assemblies (CEAs). The driving power for CEDMs is supplied by the coil stack assembly mounted on the motor housing assembly. The Coil Stack assembly is heated by the motor housing assembly contacted with reactor coolant as well as electric resistance of coil. The CEDMs are cooled by air ventilation. The cooling air around the CEDMs flows into the individual cooling shroud and extracts heat from the coil stacks while it passes between the shroud and coil housing. The cooling performance of the CEDMs is closely related to the geometry configuration of coil stack assembly. The expansion of air passage and additional cooling fins are considered to improve the cooling performance of APR+ CEDMs. In this study the optimal design conditions of APR+ CEDMs are investigated by flow analysis

  15. Sodium cooling FBR type reactor plant

    International Nuclear Information System (INIS)

    In a sodium cooling FBR type reactor plant, a steam electrolysis type hydrogen forming device using high temperature steams as the starting material and a steam turbine power generator operated by high pressure/high temperature steams are disposed. During day time when a power demand is increased, a steam switch valve on the side of the hydrogen forming device is closed and the steam switch valve on the side of the power generator is opened to introduce substantially entire amount of steams to the steam turbine. During mid night when the power demand is decreased, the steam switch valve on the side of the power generator is closed and the steam switch valve on the side of the hydrogen forming device is opened to introduce substantially entire amount of steams to the steam electrolysis device, or the exhausted gases from the steam turbine type power generator is heated again by a heat exchanger and low pressure/high temperature steams are introduced to the steam electrolysis device. In this case, electric power is applied between a hydrogen electrode and an oxygen electrode to form a hydrogen gas and an oxygen gas, and the hydrogen gas is stored in a hydrogen storage vessel. It can easily cope with the fluctuation of the power demand, as well as hydrogen can be efficiently produced. (N.H.)

  16. Transient Performance of Air-cooled Condensing Heat Exchanger in Long-term Passive Cooling System during Decay Heat Load

    International Nuclear Information System (INIS)

    In the event of a 'loss of coolant accident'(LOCA) and a non-LOCA, the secondary passive cooling system would be activated to cool the steam in a condensing heat exchanger that is immersed in an emergency cooldown tank (ECT). Currently, the capacities of these ECTs are designed to be sufficient to remove the sensible and residual heat from the reactor coolant system for 72 hours after the occurrence of an accident. After the operation of a conventional passive cooling system for an extended period, however, the water level falls as a result of the evaporation from the ECT, as steam is emitted from the open top of the tank. Therefore, the tank should be refilled regularly from an auxiliary water supply system when the system is used for more than 72 hours. Otherwise, the system would fail to dissipate heat from the condensing heat exchanger due to the loss of the cooling water. Ultimately, the functionality of the passive cooling system would be seriously compromised. As a passive means of overcoming the water depletion in the tank, Kim et al. applied for a Korean patent covering the concept of a long-term passive cooling system for an ECT even after 72 hours. This study presents transient performance of ECT with installing air-cooled condensing heat exchanger under decay heat load. The cooling capacity of an air-cooled condensing heat exchanger was evaluated to determine its practicality

  17. Design of a power conversion system for an indirect cycle, helium cooled pebble bed reactor system

    International Nuclear Information System (INIS)

    A design is presented for the turbomachinery for an indirect cycle, closed, helium cooled modular pebble bed reactor system. The design makes use of current technology and will operate with an overall efficiency of 45%. The design uses an intermediate heat exchanger which isolated the reactor cycle from the turbomachinery. This design excludes radioactive fission products from the turbomachinery. This minimizes the probability of an air ingress accident and greatly simplifies maintenance. (author)

  18. Seismic stability of VGM type high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    The main principles of the design provision of high temperature gas cooled VGM reactors seismic stability and the results of calculations, performed by linear-spectral method are presented. (author). 1 ref., 10 figs

  19. Conceptual design study of small sodium cooled reactors

    International Nuclear Information System (INIS)

    A conceptual design of various small metal fuel sodium cooled reactors has been studied in the feasibility study on commercialized fast breeder reactor cycle system. In FY2004 study, a 50 MWe power plant for remote places with a long life core without refueling and a 300 MWe modular reactor which pursues standardization for learning effect and reduction of capital risks. In the small reactor with a long life core, the reactor vessel is minimized without a permanent fuel handling system and the cooling system is simplified adopting 1 loop. The total mass of the reactor vessel and the cooling system is dramatically reduced and the concept has a potential to be an attractive power source for remote places. In the 300 MWe modular reactor, the cooling system adopts 1 loop and the ex-vessel fuel storage tank for spent fuels is eliminated adapting the in-vessel storage (IVS) which has a capacity for a 4 year storage. The reactor building is minimized without the ex-vessel storage. This concept has a potential to be an power source for key grids with modular constructions and a first plant with a small fuel cycle facility can demonstrate the metal fuel fast reactor cycle. (author)

  20. Emergency cooling system for nuclear reactors

    International Nuclear Information System (INIS)

    Upon 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. (author)

  1. Limiting Factors for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    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

  2. Startup of air-cooled condensers and dry cooling towers at low temperatures of the cooling air

    Science.gov (United States)

    Milman, O. O.; Ptakhin, A. V.; Kondratev, A. V.; Shifrin, B. A.; Yankov, G. G.

    2016-05-01

    The problems of startup and performance of air-cooled condensers (ACC) and dry cooling towers (DCT) at low cooling air temperatures are considered. Effects of the startup of the ACC at sub-zero temperatures are described. Different options of the ACC heating up are analyzed, and examples of existing technologies are presented (electric heating, heating up with hot air or steam, and internal and external heating). The use of additional heat exchanging sections, steam tracers, in the DCT design is described. The need for high power in cases of electric heating and heating up with hot air is noted. An experimental stand for research and testing of the ACC startup at low temperatures is described. The design of the three-pass ACC unit is given, and its advantages over classical single-pass design at low temperatures are listed. The formation of ice plugs inside the heat exchanging tubes during the start-up of ACC and DCT at low cooling air temperatures is analyzed. Experimental data on the effect of the steam flow rate, steam nozzle distance from the heat-exchange surface, and their orientation in space on the metal temperature were collected, and test results are analyzed. It is noted that the surface temperature at the end of the heat up is almost independent from its initial temperature. Recommendations for the safe start-up of ACCs and DCTs are given. The heating flow necessary to sufficiently heat up heat-exchange surfaces of ACCs and DCTs for the safe startup is estimated. The technology and the process of the heat up of the ACC with the heating steam external supply are described by the example of the startup of the full-scale section of the ACC at sub-zero temperatures of the cooling air, and the advantages of the proposed start-up technology are confirmed.

  3. A study of occupant cooling by personally controlled air movement

    OpenAIRE

    Arens, Edward A.; Xu, T.; Miura, K; Zhang, H.; Fountain, M.; Bauman, Fred

    1997-01-01

    This study addresses the effectiveness of air movement cooling, an alternative to compressor-based cooling of the air itself. Subjects in an environmental chamber were exposed to a range of warm temperatures and allowed to adjust air movement to suit their individual preferences, while answering a series of questions about their comfort. Air movement was from the subject's side, in two modes of turbulent flow. The air speeds chosen by the subjects, and their subjective responses, are evaluate...

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. 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.

  7. Testing stand for cosmic gas-cooling fast reactor's sample

    International Nuclear Information System (INIS)

    For carrying out of technical decision and nuclear, radiation and technological safety of gas-cooling space nuclear power plants is elaborating gas-cooling fast reactor's testing stand. In the base of its draft is taken conception of the reactor with filling up type reactor core on the base of ball fuel elements and radial coolant flowing. On the testing stand would suggested carrying out testing for study neutron and physical parameters of gas-cooling reactor, its behaviour under accident simulation. In the reactor core will suggest use carbon nitrides fuel elements with tungsten cover, provides under nominal regime relatively low fission products yield to first contour of device. Construction of fuel element was carrying out on reactor and non reactor testing and its calculated on working resource about 3000 hours. Constructive materials of reactor core have lower melting temperature, that provides organized in good time remove fuel element to containers placed under reactor in case connected with hypothetical accident. In the construction of reactor for seen tree-contours system of heat transfer and its provides multistage system of barriers against fission products yield to environment. tabs.1

  8. Efficient Water Management in Water Cooled Reactors

    International Nuclear Information System (INIS)

    number of the countries that have recently begun to consider the introduction of nuclear power are in water scarce regions, which would certainly limit the possibility for deployment of nuclear power plants, in turn hindering these countries' development and energy security. Thus, there is a large incentive to enhance efforts to introduce innovative water use, water management practices and related technologies. Water management for nuclear power plants is gaining interest in IAEA Member States as an issue of vital importance for the deployment of nuclear power. Recent experience has shown that some nuclear power plants are susceptible to prolonged drought conditions, forcing reactors to be shut down or power to be reduced to a minimal level. In some cases, environmental issues have resulted in regulations that limit the possibility for water withdrawal as well as water discharge. Regarding the most common design for cooling nuclear power plants, this has led to a complicated siting procedure for new plants and expensive retrofits for existing ones. The IAEA has already provided its Member States with reports and documents that address the issue. At the height of nuclear power expansion in the 1970s, the need for guidance in the area resulted in publications such as Thermal Discharges at Nuclear Power Stations - Their Management and Environmental Impact (Technical Reports Series No. 155) and Environmental Effects of Cooling Systems (Technical Reports Series No. 202). Today, amid the so-called nuclear renaissance, it is of vital importance to offer guidance to the Member States on the issues and possibilities that nuclear power water management brings. Management of water at nuclear power plants is an important subject during all phases of the construction, operation and maintenance of any nuclear power plant. Water management addresses the issue of securing water for condenser cooling during operation, for construction (during the flushing phase), and for inventory

  9. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    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 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  10. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    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.)

  11. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  12. The economics of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Specific technical features of fast reactors make them more expensive than water-cooled reactors in terms of initial investment, an over cost of 30% is acknowledged in this study. Their consumption of natural uranium is negligible being fed on depleted uranium (except for the very first cycle when an important quantity of plutonium is necessary). In the context of the scarcity of natural uranium, fast reactors could provide a competitive KWh compared with PWR. The study shows that sodium-cooled fast reactor could be economically competitive somewhere in the second part of the 21. century. The development of fast reactors could be accelerated by other arguments than economic competitiveness, for instance some governments might value more the energy independence given by a fleet of fast reactors or by considerations linked to non-proliferation or to the burning of actinides. In addition the article details the worldwide resource in natural uranium. (A.C.)

  13. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    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 of providing fast reactor electricity at an affordable and competitive price.

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

    International Nuclear Information System (INIS)

    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

  15. Passive Decay Heat Removal System Options for S-CO{sub 2} Cooled Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO{sub 2} Brayton power cycle. The S-CO{sub 2} power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO{sub 2} power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored.

  16. 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.

  17. Pin-Type Gas Cooled Reactor for Nuclear Electric Propulsion

    Science.gov (United States)

    Wright, Steven A.; Lipinski, Ronald J.

    2003-01-01

    This paper describes a point design for a pin-type Gas-Cooled Reactor concept that uses a fuel pin design similar to the SP100 fuel pin. The Gas-Cooled Reactor is designed to operate at 100 kWe for 7 years plus have a reduced power mode of 20% power for a duration of 5 years. The power system uses a gas-cooled, UN-fueled, pin-type reactor to heat He/Xe gas that flows directly into a recuperated Brayton system to produce electricity. Heat is rejected to space via a thermal radiator that unfolds in space. The reactor contains approximately 154 kg of 93.15 % enriched UN in 313 fuel pins. The fuel is clad with rhenium-lined Nb-1Zr. The pressures vessel and ducting are cooled by the 900 K He/Xe gas inlet flow or by thermal radiation. This permits all pressure boundaries to be made of superalloy metals rather than refractory metals, which greatly reduces the cost and development schedule required by the project. The reactor contains sufficient rhenium (a neutron poison) to make the reactor subcritical under water immersion accidents without the use of internal shutdown rods. The mass of the reactor and reflectors is about 750 kg.

  18. Safety analysis of reactor's cooling system

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  20. Filter for reactor emergency cooling system

    International Nuclear Information System (INIS)

    The invention describes the design of a filter for the emergency cooling system. The new type of filter can be rinsed by flushing water backwards through the filter. The arrangement will prevent the filter from being silt up

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

    International Nuclear Information System (INIS)

    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

  2. Containment for Heavy-Water Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    The safety principles applicable to heavy-water, gas-cooled reactors are outlined, with a view to establishing containment specifications adapted to the sites available in Switzerland for the construction of nuclear plants. These specifications are derived from dose rates considered acceptable, in the event of a serious reactor accident, for persons living near the plant, and are based on-meteorological and demographic conditions representative of the majority of the country's sites. The authors consider various designs for the containment shell, taking into account the conditions which would exist in the shell after the maximum credible accident. The following types of shell are studied: pre-stressed concrete; pre-stressed concrete with steel dome; pre-stressed concrete with inner, leakproof steel lining; steel with concrete side shield to protect against radiation; double shell. The degree of leak proofing of the shells studied is regarded as a feature of the particular design and not as a fixed constructional specification. The authors assess the leak proofing properties of each type of shell and establish building costs for each of them on the basis of precise plans, with the collaboration of various specialized firms. They estimate the effectiveness of the various shells from a safety standpoint, in relation to different emergency procedures, in particular release into the atmosphere through appropriate filters and decontamination of the air within the shell by recycling through batteries of filters. The paper contains a very detailed comparison of about 10 cases corresponding to various combinations of design and emergency procedure; the comparison was made using a computer programme specially established for the purpose. The results are compared with those for a reactor of the same type and power, but assembled together with the heat exchangers in a pre-stressed concrete shell. (author)

  3. Gas-cooled reactor application for a university campus

    International Nuclear Information System (INIS)

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

  4. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    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

  5. Impact of inlet air cooling on gas turbine performance

    OpenAIRE

    Szymon Jarzębowski; Ewa Pyzik; Andrzej Miller

    2012-01-01

    In this article different possibilities of gas turbine inlet air cooling were presented. The method of defining power gain caused by air cooling was discussed. The results of increasing power output level of several different turbines and one gas turbine in combine cycle in domestic ambient conditions were presented and discussed. Significant turbine power gains were received.

  6. Materials development and materials selection for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    As applied to operational conditions of fast reactors a combined investigations of structural materials are accomplished. Steels 10Kh18N9, 08Kh16N11M3 are recommended to be used for reactor vessels, steel 10Kh2M - for steam generators, a strain hardened steel 08Kh16N11M3T and steel 05Kh12N2M - for fuel assembly cans. The investigations provided for designing such sodium cooled fast reactors as Bor-60, BN-350, BN-600. The investigation results are now in use in construction of a new fast reactor BN-800

  7. A preliminary study on HTGR with air-cooled condenser at Riyadh, Saudi Arabia

    International Nuclear Information System (INIS)

    N GNP reactor plant adopted a Rank in steam cycle for early deployment and for reducing R and D risk and cost. Original plant design is based on a wet cooling tower with wet bulb temperature of 34 .deg. C. This cooling environment may be sufficient for most area in North America. However, we should consider air temperature of 45 .deg. C and no available cooling water for any site near Riyadh, Saudi Arabia. A plausible option in such arid area is using an air-cooled condenser(ACC) which is widely used in a combined cycle plant in arid region. ACC is also suitable for freezing area such as northern territory or high elevation remote area. We have studied impact of the cooling method on the power generation efficiency and the annual average power production referencing N GNP steam turbine. Even though condenser split is assumed to be the same between ACC and wet cooling tower, large difference in air temperature and wet bulb temperature makes large efficiency loss in the ACC. The ACC efficiency is lower than that of the wet cooling tower by 1.12%. To make up this loss, we proposed the variable steam extraction rates operation. An air cooled condenser is a practical

  8. Design Study of Air Cooled GTHTR300A for inland installation

    International Nuclear Information System (INIS)

    GTHTR300 is a design series of power and cogeneration plants based on common technologies of a high temperature gas reactor and a direct Brayton cycle gas turbine power generator. Although the series currently selects seawater cooling as the method to reject reactor waste, heat, the employment of the Brayton cycle offers potential conditions for economical dry air cooling This paper reports the results of a conceptual design study for a new plant entry in the series, namely GTHTR300A for air cooling. Dry air cooling is made to replace seawater cooling of the baseline plant without forcing major redesign and new development in the reactor or in the balance of plant as is done with previous baseline derivatives. For example, the core layout and the reactor pressure vessel dimensions are not changed from the baseline even if their optimization may produce small cost saving. It is also found that by down scaling reactor thermal power by 20%, the size for the balance of plant is unchanged while delivering the same level of plant generating efficiency. An original design feature introduced in GTHTR300A is use of an intermediate helium loop, instead of usually water, to transport waste heat from reactor to cooling tower. Water is not used so that the plant can rule out accident of water ingress into reactor. By operating at high pressure but low temperature, circulating power consumption and equipment cost are kept down for the loop. The air cooling tower employs natural draft although mechanical draft, not detailed in this paper, has also been found acceptable to this plant. A compact tower is obtained due to the high thermal efficiency of 48.2%, as opposed to 34.5% of today's nuclear plants, and to the fact that the cycle rejects waste heat in a high temperature range. The maximum design sharing, high thermal efficiency, enhanced safety, and economical dry cooling make the GTHTR300A practical for inland installation without reliance on a large source of water to reject

  9. 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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  12. Cooling Performance of ALIP according to the Air or Sodium Cooling Type

    International Nuclear Information System (INIS)

    ALIP pumps the liquid sodium by Lorentz force produced by the interaction of induced current in the liquid metal and their associated magnetic field. Even though the efficiency of the ALIP is very low compared to conventional mechanical pumps, it is very useful due to the absence of moving parts, low noise and vibration level, simplicity of flow rate regulation and maintenance, and high temperature operation capability. Problems in utilization of ALIP concern a countermeasure for elevation of internal temperature of the coil due to joule heating and how to increase magnetic flux density of Na channel gap. The conventional ALIP usually used cooling methods by circulating the air or water. On the other hand, GE-Toshiba developed a double stator pump adopting the sodium-immersed self-cooled type, and it recovered the heat loss in sodium. Therefore, the station load factor of the plant could be reduced. In this study, the cooling performance with cooling types of ALIP is analyzed. We developed thermal analysis models to evaluate the cooling performance of air or sodium cooling type of ALIP. The cooling performance is analyzed for operating parameters and evaluated with cooling type. 1-D and 3-D thermal analysis model for IHTS ALIP was developed, and the cooling performance was analyzed for air or sodium cooling type. The cooling performance for air cooling type was better than sodium cooling type at higher air velocity than 0.2 m/s. Also, the air temperature of below 270 .deg. demonstrated the better cooling performance as compared to sodium

  13. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)

    2015-05-15

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.

  14. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity

  15. New reactor cavity cooling system using novel shape for HTGRs and VHTRs

    International Nuclear Information System (INIS)

    A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal. (author)

  16. Development of cooling strategy for an air cooled lithium-ion battery pack

    Science.gov (United States)

    Sun, Hongguang; Dixon, Regan

    2014-12-01

    This paper describes a cooling strategy development method for an air cooled battery pack with lithium-ion pouch cells used in a hybrid electric vehicle (HEV). The challenges associated with the temperature uniformity across the battery pack, the temperature uniformity within each individual lithium-ion pouch cell, and the cooling efficiency of the battery pack are addressed. Initially, a three-dimensional battery pack thermal model developed based on simplified electrode theory is correlated to physical test data. An analytical design of experiments (DOE) approach using Optimal Latin-hypercube technique is then developed by incorporating a DOE design model, the correlated battery pack thermal model, and a morphing model. Analytical DOE studies are performed to examine the effects of cooling strategies including geometries of the cooling duct, cooling channel, cooling plate, and corrugation on battery pack thermal behavior and to identify the design concept of an air cooled battery pack to maximize its durability and its driving range.

  17. Evaluation for External Reactor Vessel Cooling System using CFD Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Seok Bin; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2012-05-15

    To ensure the safety of the nuclear plants, there are lots of safety systems in the nuclear plant. One of them is External Reactor Vessel Cooling system (ERVC) which is operated when a molten corium is relocated in a lower head of a reactor vessel. As ERVC system runs, coolant flows down into a reactor cavity to remove a decay heat from the molten corium. This work simulated the ERVC system which is applied to APR1400 with CFD. To estimate the efficiency of the ERVC system, we designed the reactor cavity of the ERVC system of APR1400 in a full scale. From the designed model, we measured temperature distribution of the reactor vessel outer wall. Two kinds of coolant were used in this computational approach. One is present flooding matter which is water. The other is liquid metal gallium. With varying the area of the inlet and outlet of reactor cavity, we evaluated the importance of each variable

  18. Preliminary evaluation of alternate-fueled gas cooled fast reactors

    International Nuclear Information System (INIS)

    A preliminary evaluation of various alternative fuel cycles for the Gas-Cooled Fast Reactor (GCFR) is presented. Both homogeneous and heterogeneous oxide-fueled GCFRs are considered. The scenario considered is the energy center/dispersed reactor concept in which proliferation-resistant denatured reactors are coupled to 233U production reactors operating in secure energy centers. Individual reactor performance characteristics and symbiotic system parameters are summarized for several possible alternative fuel concepts. Comparisons are made between the classical homogeneous GCFR and the advanced heterogeneous concept on the basis of breeding ratio, doubling time, and net fissile gain. In addition, comparisons are made between a three-dimensional reactor model and the R-Z heterogeneous configuration utilized for the depletion and fuel management calculations. Lastly, thirty-year mass balance data are given for the various GCFR fuel cycles studied

  19. Feasibility and deployment strategy of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    The author have studied water cooled thorium breeder reactor based on matured pressurized water reactor (PWR) plant technology for several years. Through these studies it is concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array with heavy water coolant in the primary loop by replacing original light water is appropriate for achieving sufficient breeding performance as sustainable fission system and high enough burn-up as an economical power plant. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste

  20. The case for the gas cooled fast reactor

    International Nuclear Information System (INIS)

    Although gas-cooling for fast reactors had been the subject of consideration since the early days of nuclear power, it was when the concept of the prestressed concrete pressure vessel turned into practical fact, that convincing arguments could be made to overcome safety objections. In terms of hardware, the Gas Cooled Fast Breeder Reactor can rely on existing and available technologies; as far as fuel is concerned, valuable information will be derived from the Liquid Metal Fast Breeder Reactor programme. The GCFR can be made very flexible; its capital cost will not exceed by more than 20% the one for reactor built at present on commercial scale; the overall economy of its fuel cycle is good. It could play an important role in the future breeder family

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

    International Nuclear Information System (INIS)

    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

  2. PARAMETRIC SENSITIVITY STUDY OF A FUSION REACTOR DIVERTOR COOLING FINGER

    OpenAIRE

    Martin, Oliver; SIMONOVSKI IGOR

    2012-01-01

    In this paper the results of coupled thermal-mechanical Finite Element (FE) analysis on the design of a fusion reactor divertor cooling finger are presented. Beside its main purpose, to remove alpha particles, helium and other impurities from the plasma stream, a divertor has to remove approximately 15% of the total thermal power of the fusion reactor. The aim of the analysis is to assess the influence of a number of physical properties of the brazing layer (BL) of the cooling finger on the o...

  3. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    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

  4. Indirect Evaporative Pre-Cooled Compressor Cooling System Performance under Various Outdoor Air Humidity Conditions

    OpenAIRE

    Brahmanis, A; Lešinskis, A

    2013-01-01

    The present study is devoted to efficiency evaluation of a combined indirect evaporative – compressor cooling system under various outdoor air humidity conditions of temperate climate. The investigated system is located in the recently restored historical building, The Art Museum Riga Bourse, which was initially built in the middle of the 19th century. The indirect adiabatic chiller supplies cooled fluid to the conventional cooling system, consisting of ventilation cooling coils and fan-coil ...

  5. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

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

  6. Capital cost: gas cooled fast reactor plant

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-01

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

  8. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    • 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

  9. Air cooled turbine component having an internal filtration system

    Science.gov (United States)

    Beeck, Alexander R.

    2012-05-15

    A centrifugal particle separator is provided for removing particles such as microscopic dirt or dust particles from the compressed cooling air prior to reaching and cooling the turbine blades or turbine vanes of a turbine engine. The centrifugal particle separator structure has a substantially cylindrical body with an inlet arranged on a periphery of the substantially cylindrical body. Cooling air enters centrifugal particle separator through the separator inlet port having a linear velocity. When the cooling air impinges the substantially cylindrical body, the linear velocity is transformed into a rotational velocity, separating microscopic particles from the cooling air. Microscopic dust particles exit the centrifugal particle separator through a conical outlet and returned to a working medium.

  10. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  11. Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    C. Davis; S. Herring; P. MacDonald; K. McCarthy; V. Shah; K. Weaver (INEEL); J. Buongiorno; R. Ballinger; K. Doyoung; M. Driscoll; P. Hejzler; M. Kazimi; N. Todreas (MIT)

    1999-07-01

    The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.

  12. Cooling of concrete structure in advanced heavy water reactor

    International Nuclear Information System (INIS)

    Innovative nuclear power plants are being designed by incorporation of passive systems to the extent possible for enhancing the safety by elimination of active components. BARC has designed Advanced Heavy Water Reactor (AHWR) incorporating several passive systems to facilitate the fulfillment of safety functions of the reactor during normal operation, residual heat removal, emergency core cooling, confinement of radioactivity etc. In addition to these passive systems, an innovative passive technology is being developed to protect, the concrete structure in high temperature zone (V1-volume). Passive Concrete Cooling System (PConCS) uses the principle of natural circulation to provide cooling outside the insulation cabinet encompassing high temperature piping. Cooling water is circulated from overhead GDWP in cooling pipes fixed over corrugated plate on outer surface of insulation cabinet and maintains low temperature of concrete structure. Modular construction of insulation cabinet and cooling pipes external to the concrete surface simplifies the design, construction and refurbishment if required. The paper describes the details of passive technology for concrete cooling. (author)

  13. Gamma spectroscopy in water cooled reactors

    International Nuclear Information System (INIS)

    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

  14. Thermal Sizing of Heat Exchanger Tubes for Air Natural Convective Cooling System of Emergency Cooling Tank

    International Nuclear Information System (INIS)

    For the long operation of secondary passive cooling system, however, water level goes down by evaporation in succession at emergency cooling tank. At the end there would be no place to dissipate heat from condensation heat exchanger. Therefore, steam cooling heat exchanger is put on the top of emergency cooling tank to maintain appropriate water level by collecting evaporating steam. Steam cooling heat exchanger is installed inside an air chimney and evaporated steam is cooled down by air natural convection. In this study, thermal sizing of steam cooling heat exchanger under air natural convection was conducted by TSCON program for the design of experimental setup as shown in Fig. 2. Thermal sizing of steam cooling heat exchanger tube under air natural convection was conducted by TSCON program for the design of experimental setup. 25 - 1' tubes which has a length 1687 mm was determined as steam cooling heat exchanger at 2 kW heat load and 100 liter water pool in emergency cooling tank (experimental limit condition). The corresponding width of two tubes is 50 mm and has 5 by 5 tube array for heat exchanger

  15. A passive decay heat removal system for LWRs based on air cooling

    International Nuclear Information System (INIS)

    Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate

  16. A passive decay heat removal system for LWRs based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2015-05-15

    Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

  17. Air-cooled CWS warm air furnace. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Litka, A.F.; Becker, F.E.

    1995-08-01

    Thermo Power Corporation, Tecogen Division, has developed coal water slurry (CWS) combustion technologies specifically tailored to meet the space heating needs of the residential, commercial, and industrial market sectors. This furnace was extensively tested and met all the design and operating criteria of the development program, which included combustion efficiencies in excess of 99%, response to full load from a cold start in less than 5 minutes, and steady-state thermal efficiencies as high as 85%. While this furnace design is extremely versatile, versatility came at the expense of system complexity and cost. To provide a more cost effective CWS-based option for the residential market sector, Tecogen, developed a totally air-cooled CWS-fired residential warm air heating system. To minimize system cost and to take advantage of industry manufacturing practices and experience, a commercially available oil/gas solid fuel-fired central furnace, manufactured by Yukon Energy Corporation, was used as the platform for the CWS combustor and related equipment. A prototype furnace was designed, built, and tested in the laboratory to verify system integrity and operation. This unit was then shipped to the PETC to undergo demonstration operation and serve as a showcase of the CWS technology. An in-depth Owners Manual was prepared and delivered with the furnace. This Owners Manual, which is included as Appendix A of this report, includes installation instructions, operating procedures, wiring diagrams, and equipment bulletins on the major components. It also contains coal water slurry fuel specifications and typical system operating variables, including key temperatures, pressures, and flowrates.

  18. Emergency core cooling during an SRS reactor LOPA

    International Nuclear Information System (INIS)

    The loss-of-pumping accident (LOPA) is a Savannah River site (SRS) reactor design-basis accident. The most limiting LOPA is caused by a double-ended guillotine break in a secondary cooling system inlet header and is the topic of this discussion. Upon break detection, the reactor scrams and the secondary cooling water pumps and alternating-current (ac) primary pump motors trip off. The direct-current (dc) motors continue to drive the primary pumps at about one-third capacity. Gravity flow through the broken header continues flooding the building after the cooling pumps are off. The emergency cooling system (ECS) is activated prior to flood-out of the dc motors. The design-basis accident reactor power limit ensures the reactor will shut down safely should a LOPA occur. The simulated LOPA has five phases: steady state, ac coastdown, dc flow, dc coastdown, and fully developed ECS flow. Analyses of LOPAs have shown that ECS is the most limiting phase of the accident. This paper concentrates on the role of ECS in LOPA limits

  19. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

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

  20. IAEA activities in Gas-cooled Reactor technology development

    International Nuclear Information System (INIS)

    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

  1. IAEA activities in gas-cooled reactor technology development

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  3. Three decades of experience with cooling water system of a fast reactor

    International Nuclear Information System (INIS)

    The cooling water system constitutes the terminal heat exchange system for the fast breeder test reactor (FBTR) which is a sodium cooled fast reactor of 40 MWt capacity. It transfers the residual heat to atmosphere through a cooling tower. Cooling water system of FBTR comprises two sub-systems namely condenser cooling water system and service water system. Condenser cooling water is circulated through main condenser, dump condenser, condensate cooler, generator air cooler and turbine oil cooler. Service water system removes heat from several heat exchangers of auxiliary systems like air compressor, cold trap cooling, nitrogen plant, Biological Shield Cooling (BSC), Diesel Generator (DG) and steam-water system sample coolers. The cooling water system consists of an open recirculating type with an induced draft cooling tower as the ultimate heat sink. Initially, Palar river water was used as the cooling medium. At present, due to scarcity of river water, sub soil water and output from Nuclear Desalination Demonstration Plant (NDDP) are also used as cooling water. The material of construction of pipe line is carbon steel and the heat exchanger tube and other equipment materials are copper, admiralty brass, aluminium brass, bronze, Cu-Ni and carbon steel. The construction of the cooling water system of FBTR was completed in 1980. Since then the sub-systems were commissioned one by one. Whenever a sub system was commissioned, it generated a lot of impurities which affected the existing treatment programme. Sodium hexa meta phosphate treatment, Langelier Index monitoring, chlorination, global and target dispersant addition at high heat flux heat exchanger, chemical cleaning of corroded pipelines, corrosion monitoring, side stream filtration, addition of phosphonate-based corrosion inhibitor, broad spectrum biocide and specific biocide for iron oxidising bacteria are some of the phases of the cooling water treatment programme. At present, corrosion rates are generally

  4. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    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. Consequences of reactor fuel damage: - Production of radioactive wastes. - Radioactivity in the reactor cooling system

    International Nuclear Information System (INIS)

    The report describes the consequences of damage of reactor fuel cladding. The types of damage and the release of fission products into the reactor cooling system are described as well as detection methods. The report also gives suggestions to reduce the consequences of a damage. (62 figs., 13 tabs.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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. 7600C (14000F) 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

  8. 21 CFR 211.46 - Ventilation, air filtration, air heating and cooling.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 4 2010-04-01 2010-04-01 false Ventilation, air filtration, air heating and... Buildings and Facilities § 211.46 Ventilation, air filtration, air heating and cooling. (a) Adequate... holding of a drug product. (c) Air filtration systems, including prefilters and particulate matter...

  9. Storage of HLW in engineered structures: air-cooled and water-cooled concepts

    International Nuclear Information System (INIS)

    A comparative study on an air-cooled and a water-cooled intermediate storage of vitrified, highly radioactive waste (HLW) in overground installations has been performed by Nukem and Belgonucleaire respectively. In the air-cooled storage concept the decay heat from the storage area will be removed using natural convection. In the water-cooled storage concept the decay heat is carried off by a primary and secondary forced-cooling system with redundant and diverse devices. The safety study carried out by Nukem used a fault tree method. It shows that the reliability of the designed water-cooled system is very high and comparable to the inherent, safe, air-cooled system. The impact for both concepts on the environment is determined by the release route, but even during accident conditions the release is far below permissible limits. The economic analysis carried out by Belgonucleaire shows that the construction costs for both systems do not differ very much, but the operation and maintenance costs for the water-cooled facility are higher than for the air cooled facility. The result of the safety and economic analysis and the discussions with the members of the working group have shown some possible significant modifications for both systems, which are included in this report. The whole study has been carried out using certain national criteria which, in certain Member States at least, would lead to a higher standard of safety than can be justified on any social, political or economic grounds

  10. High temperature gas-cooled reactors - perspective of thermal reactor concept with high thermal efficiency

    International Nuclear Information System (INIS)

    The present HTR development is based worldwide on the extensive experience gained in the construction and operation of gas-cooled reactors of the Magnox type and on the successful operation of the experimental high temperature reactors Dragon, Peach Bottom and AVR. The advanced CO2-cooled reactors, as well as the HTR prototype power plants for St. Vrain and THTR, are all suffering considerable delays in construction and commissioning. The commercial HTR plants have not yet achieved the decisive breakthrough onto the market. Increasing interest is being shown in advanced HTR systems, i.e., HTR with gas turbine, HTR process heat reactors and gas-cooled fast breeders. The key problem in the coming years will be the closing of the fuel cycle. Development work in this connection has already started. (orig.)

  11. Creation of Air-Cooled Mn Series Bainitic Steels

    Institute of Scientific and Technical Information of China (English)

    FANG Hong-sheng; FENG Chun; ZHENG Yan-kang; YANG Zhi-gang; BAI Bing-zhe

    2008-01-01

    The development and mechanical performances of new type air-cooled Mn series bainitie steels including granular bainitie steels,FGBA/BG duplex steels,CFB/M duplex steels,medium carbon bainite/martensite steels,cast hainitic steels invented by the authors are summarized.The novel series of bainitie steels are alloyed with Mn,and several series bainitic duplex microstructures can be easily obtained under the condition of air cooling through unique composition design.The invented idea,the principle of alloying design,the strengthening mechanism,and the evolution of the microstructure of new type air-cooled Mn series bainitic steels are presented.Furthermore,the applications in different fields of these Mn series air-cooled bainitic steels with different strength level are also introdueed.It is suggested that the significance of the development of the air-cooled Mn series bainitic steel can be summarized as follows:reducing costs of both raw materials and production;good combination of strength and toughness;self-hardening with high bainitic hardenability by air cooling from hot working without additional quenching-tempering treatment or quenching procedure;large savings in energy resources;and reduced environmental pollution.

  12. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    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

  13. Heat Transfer Distribution for Reactor Cavity Cooling System Riser Considering Thermal Conduction

    International Nuclear Information System (INIS)

    The reactor cavity cooling system (RCCS) is the safety grade system for a very high temperature reactor (VHTR). The main role of the RCCS is the heat removal from the reactor vessel. The verification of the performance of the passive RCCS is a key objective for the construction of demonstration plant. Korea Atomic Energy Research Institute (KAERI) is considering the air-cooled RCCS under natural convection operation for a PMR200. Bae et al. studied scaling of PMR200 RCCS prior to the experimental verification of the RCCS. The cavity radiation number and temperature ratio number were selected as controlling non-dimensional group. The heat transfer in the riser is one of the key phenomena to predict the performance of the prototype RCCS. The riser absorbs heat from the reactor vessel, through radiative heat transfer, and from the reactor cavity, through convective heat transfer. The absorbed heat is removed by convective heat transfer in the riser. CFD results showed that thermal conduction through the RCCS riser was not negligible. Because the material of the riser duct was carbon steel, the thermal conduction effect was comparable even with a small duct thickness to the convective heat transfer by air flow. It is strongly recommended that the heat transfer in RCCS should go with consideration on the thermal conduction of the RCCS riser duct

  14. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    OpenAIRE

    Phani Kumar Domalapally; Slavomir Entler

    2015-01-01

    Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  15. Safety concept of lead-cooled fast reactors

    International Nuclear Information System (INIS)

    design characteristics of BREST-OD-300. Special emphasis is placed on demonstration of safety. The paper identifies the ways and means of attaining natural safety, which exclude prompt neutron excursion, loss of cooling, and fuel failure owing to the small reactivity margin, Δρtot ∼ β, low pressure in the circuit, large margins to limiting temperatures, high level of natural circulation, passive removal of decay heat by air (unlimited in time), high heat storage capacity of the lead circuit, passive coolant temperature and flow rate feedbacks, etc. The BREST reactor safety was verified by successive analyses of accidents arising from the most hazardous and credible initiating events and by proving deterministically the impossibility of the off-normal conditions growing into an accident which would require evacuation of local residents. The only features to ensure reactor stability considered in the accident analyses were the feedbacks of core components and coolant flow rates. In some severe cases, consideration was also given to operation of threshold passive safety features responding to coolant flow drop and excess of the limiting temperature at the core outlet. (author)

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

    International Nuclear Information System (INIS)

    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.

  17. Studies on air ingress for pebble bed reactors

    International Nuclear Information System (INIS)

    A loss-of-coolant accident (LOCA) has been considered a critical event for helium-cooled pebbled bed reactors. Following helium depressurization, it is anticipated that unless countermeasures are taken air will enter the core through the break and then by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure and graphite pebbles. Thus, without any mitigating features a LOCA will lead to an air ingress event. The INEEL is studying such an event with two well-respected light water reactor transient response codes: RELAP5/ATHENA and MELCOR. To study the degree of graphite oxidation occurring due to an air ingress event, a MELCOR model of a reference pebble bed design was constructed. A modified version of MELCOR developed at INEEL, which includes graphite oxidation capabilities, and molecular diffusion of air into helium was used for these calculations. Results show that the lower reflector graphite consumes all of the oxygen before reaching the core. The results also show a long time delay between the time that the depressurization phase of the accident is over and the time that natural circulation air through the core occurs. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    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

  20. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

  1. The new cooling system of the IPR-RI reactor

    International Nuclear Information System (INIS)

    A 250 Kw cooling system was designed and installed to cool the demineralized water of the IPR-RI reactor (Triga Mark I). The primary circuit is made of 304 stainless steel. The secondary circuit is made of carbon steel, and the cooling tower of fiberglass.The heat exchanger is of the shell-and-tube type with four tube passes. The demineralized water (primary side) flows through the shell and the ordinary water flows inside the tubes. This arrangement allows cleaning of the tubes (straight tubes) by simply disassembling the top of the heat exchanger. The cooling system was designed to operate at a pool temperature of 40.7 deg C. Cold water returns to the reactor tank in summer days at a maximum temperature of 33.1 deg C. The water flow in the primary side can be varied from 28 m3/h to 32 m3/h. Interlocks and protection devices coupled to pressure gauges, temperature and water level meters provide a very safe and reliable performance to the cooling system. (author)

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

    International Nuclear Information System (INIS)

    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 H2 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

  3. Experimental studies on natural circulation decay heat removal in Japan Sodium Cooled Fast Reactor (JSFR)

    International Nuclear Information System (INIS)

    Fully natural circulation system is adopted in a decay heat removal system (DHRS) of the designs of Japan Sodium Cooled Fast Reactor (JSFR). Several investigations of experiments and simulation methods on this DHRS were performed. Water experiments were carried out for the primary heat transportation system including a reactor vessel and heat exchangers of DHRS using a 1/10 model. As for the DHRS loop, sodium experiments were carried out, especially for a heat exchanger installed in an Intermediate Heat Exchanger (IHX). Here, several results of the sodium experiments were described. Transient characteristics during the start up in the air system of the air cooler, secondary loop of DHRS, and the primary loop were examined by the sodium experiments. Smooth increases of natural circulation flow rates in all systems of air and sodium were confirmed. Verifications of numerical simulation methods are planned based on the water and sodium experiments in this investigation plan. (author)

  4. Liquid-cooled nuclear reactor, especially a boiling water reactor

    International Nuclear Information System (INIS)

    A nuclear reactor with a special arrangement of fuel rods in the core is designed. Each fuel element has its shaft which is made of sheets, has the same cross section as the fuel element and protrudes at least the length of the control rod above the reactor core. Made of a zirconium alloy in the core area and of stainless steel above it, the shaft is equipped with channels for sliding the rods in and out and serves to spatially secure the position of the rods. Coolant flow is provided by the chimney effect. The shaft can conveniently enclose the control rod drive. It can also serve to bear the water separator. Moreover, it can constitute a part of the casing which surrounds the fuel rods and keeps the fuel in an intimate contact with the coolant; the other part of this casing is constituted by inserted sheets which can conveniently have the shape of angles. The walls of neighboring shafts form a compartment accommodating a neutron absorber plate. (M.D.). 11 figs

  5. Air cooling effect of fins on a Honda shine bike

    OpenAIRE

    Padhiyar Abhesinh J; Vasim G Machhar

    2015-01-01

    The main of aim of this work is to study various researches done in past to improve heat transfer rate of cooling fins by changing cylinder block fin geometry. Low rate of heat transfer through cooling fins is the main problem in this type of cooling. So efficiency of the engine is increase by increase the heat transfer. Examples of direct air cooling in modern automobiles are rare. The most common example is the commercials Automobile bike like a Honda Shine, Bajaj bike, Honda sp...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-06

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

  7. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  8. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  10. Varying duty operation of air-cooled condenser units

    Science.gov (United States)

    Milman, O. O.; Kondratev, A. V.; Ptakhin, A. V.; Dunaev, S. N.; Kirjukhin, A. V.

    2016-05-01

    Results of experimental investigations of operation modes of air-cooled condensers (ACC) under design and varying duty conditions are presented. ACCs with varying cooling airflow rates under constant heat load and with constant cooling airflow under varying heat load are examined. Diagrams of heat transfer coefficients and condensation pressures on the heat load and cooling airflow are obtained. It is found that, if the relative heat load is in the range from 0.6 to 1.0 of the nominal value, the ACC heat transfer coefficient varies insignificantly, unlike that of the water-cooled surface condensers. The results of the determination of "zero points" are given, i.e., the attainable pressure in air-cooled condensing units (ACCU), if there is no heat load for several values of working water temperature at the input of water-jet ejectors and liquid ring vacuum pump. The results of the experimental determination of atmospheric air suction into the ACC vacuum system. The effect of additional air suctions in the steam pipe on ACCU characteristics is analyzed. The thermal mapping of ACC heat exchange surfaces from the cooling air inlet is carried out. The dependence of the inefficient heat exchange zone on the additional air suction into the ACC vacuum system is given. It is shown that, if there is no additional air suction into the ACC vacuum system, the inefficient heat exchange zone is not located at the bottom of the first pass tubes, and their portion adjacent to the bottom steam pipe works efficiently. Design procedures for the ACC varying duty of capacitors are presented, and their adequacy for the ACCU varying duty estimation is analyzed.

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

    International Nuclear Information System (INIS)

    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

  12. Decay heat removal in GEN IV gas cooled fast reactors

    International Nuclear Information System (INIS)

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

  13. Economic competitiveness requirements for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    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)

  14. 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.

  15. Experimental Analysis Of 80 Tr Capacity Air Cooled Scroll Chiller Using R-22 & R-407c.

    OpenAIRE

    Mr. Bhikhu B,; Prof. Ronak Shah

    2014-01-01

    In air conditioning systems, chilled water is typically distributed to heat exchangers, or coils, in air handling units or other types of terminal devices which cool the air in their respective space(s), and then the water is recirculated back to the chiller to be cooled again. These cooling coils transfer sensible heat and latent heat from the air to the chilled water, thus cooling and usually dehumidifying the air stream. The experiment works on 80 TR capacity Air Cooled Scroll ...

  16. Status of advanced technology and design for water cooled reactors: Heavy water reactors

    International Nuclear Information System (INIS)

    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

  17. Liner insulation for gas cooled reactors - a personal history

    International Nuclear Information System (INIS)

    The paper describes briefly the development work leading to the foil and mesh insulation used at Oldbury and the fibre insulation used in the Hinkley and Hunterston reactors. The subject is covered under the headings: cooling system; insulation specification; thermal problems; Oldbury (choice of insulant; conductivity tests; commissioning); Hinkley (choice of insulant; thermal performance; mechanical details; acoustic testing; thermal cycling tests); commissioning tests; Torness and Heysham. (U.K.)

  18. The status of graphite development for gas cooled reactors

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  2. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    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

  3. Safety analysis of spent fuel element storage in 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Approximately 90000 spent fuel elements will be discharged from a 10 MW high temperature gas-cooled reactor (HTR-10) in its lifetime. The activity of the radioactive fission products in these spent fuel elements will reach 1.0 x 1016 Bq, so these spent fuel elements should be properly managed. HTR-10 spent fuel elements will be discharged into lead-steel containers, with each container designed to receive 2000 fuel elements. These containers will be stored in a concrete compartment inside the reactor building and cooled by air. The author analyzes the release of the radioactive nuclides, the critical safety parameters and the irradiation shielding. The results show that the safety requirements can be met in the HTR-10 spent fuel element storage compartment

  4. Analysis of sodium-cooled fast reactor operations world-wide and consequences for future reactor design and operation

    International Nuclear Information System (INIS)

    In 2007, the sodium-cooled fast reactors which are currently operating or have operated throughout the world will have accumulated collective experience of 379 years of operation for 18 different reactors. This paper summarizes the various incidents and problems which have impacted these reactors' operations, ranks them by function and lists the solutions which were brought. This paper solely looks at the operating problems which occurred in the 'sodium' part of the fast reactors. Problems encountered in the classic water/steam part are not described herein review is made for all these reactors on: The water/sodium reactions occurred in the steam generators; -The technical difficulties on the primary components (pumps and exchangers) and the repair operations; -The incidents in handling operations; - The spurious leaks or transfers of sodium and their consequences; - The intakes of air or impurities; -The experience from fuel and clad failures; - The neutronic operations and control; -The material behaviour problems (as 321 or SPX1 drum); - The difficulties due to sodium aerosols. And all the specific difficulties due to sodium related technological problems. This paper also describes how the experience gained has been taken into account in the safe operation of these plants and also in the design of future reactors. n conclusion it appears that a significant experience has been accumulated that allow today good availability for the remaining operating plants and also good design possibilities for the future. (author)

  5. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

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

    International Nuclear Information System (INIS)

    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

  7. Cooling of core debris within the reactor vessel lower head

    International Nuclear Information System (INIS)

    Under severe-accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. The Three Mile Island Unit 2 (TMI-2) accident demonstrated that this could be accomplished by water resident within the reactor vessel combined with injection on a continual basis to quench the debris and remove decay heat over the long term. Some accident situations could result in the transport of molten core debris to the lower plenum, as occurred in TMI-2, the boil-off of water in the lower plenum, and an inability to add water to the reactor coolant system (RCS). Even in this extreme set of circumstances, sufficient cooling may be available to prevent failure of the reactor pressure vessel (RPV) lower head and thereby retain the core debris within the vessel. Experiments were performed in support of Commonwealth Edison's Zion individual plant examination and accident management programs that demonstrate nucleate boiling heat removal rates from the outer surface of a simulated RPV lower head surrounded by typical reflective insulation used in nuclear power plants

  8. Cooling of core debris within the reactor vessel lower head

    International Nuclear Information System (INIS)

    Under severe accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. Some accident situations could result in the transport of molten core debris to the lower plenum, as occurred to some extent (∼20 tonnes) during the TMI-2 accident, boiloff of water in the lower plenum, and an inability to add water to the reactor coolant system (RCS). In this extreme set of circumstances, sufficient external reactor pressure vessel (RPV) cooling may be available to prevent failure of the RPV lower head and, thereby, retain the core debris within the vessel. Containment configurations like Zion would result in substantial accumulation of water around the lower parts of the reactor vessel for most accident sequences. The experiments which were performed in support of the Commonwealth Edison individual plant examination and accident management programs, are heat transfer tests designed to demonstrate that nucleate boiling is the dominant heat removal process from the outer surface of a simulated RPV lower head surrounded by typical reflective insulation used in nuclear power plants

  9. Thermophysical properties of materials for water cooled reactors

    International Nuclear Information System (INIS)

    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

  10. Modification of the Core Cooling System of TRIGA 2000 Reactor

    Science.gov (United States)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

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

    International Nuclear Information System (INIS)

    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 (CO2) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  12. Hybrid simulation of high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    A hybrid simulator was made to calculate the dynamics of high temperature gas cooled reactor(VHTR). The continuous space-discrete time (CSDT) method is applied to solve the partial differential equations of the heat transfer in the hybrid computation. By this method the error of the heat balance is decreased to less than one percent in the steady state. Though the mini computer is used for this simulator, it operates about five times faster than real time. The dynamics of VHTR are characterized by the large heat capacity of the reactor core and the long time constant. The values of these parameters are reported as the results of this calculation. The control system of the reactivity and the coolant flow rate is required to operate the reactor. The nonlinearity of VHTR which occurs in the change of flow rate are also understood quantitatively by this simulator. (author)

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

    International Nuclear Information System (INIS)

    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

  14. Fort St. Vrain high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The construction, testing, and preliminary operating experience of the Fort St. Vrain Nuclear Generating Station are described. This station utilizes the advanced high-temperature gas-cooled reactor (HTGR) concept and is the first nuclear reactor system in the United States to use a prestressed concrete reactor vessel (PCRV). Helium is used as the primary coolant, and a nitrogen system provides refrigeration for the low temperature equipment of the helium purification system and for the moisture monitors in the primary coolant system. Design, construction and testing to date at this station have made a significant contribution to the HTGR concept for central station electric generating plants to supply the increasing demands for electrical energy. (U.S.)

  15. Simulation on primary coolant system of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    In this paper the thermal-hydraulic characteristics of the primary loop of China Experimental Fast Reactor (CEFR) are calculated and analyzed. A one-dimension, single-phase flow model is used to establish the system control equations. The single channel model is adopted in the reactor core, and a dynamic model of intermediate heat exchanger is built. At the same time, the property of sodium and flow and heat transfer correlations or models of sodium are collected and compiled. The discussion of the sensitivity of different flow and heat transfer correlations is given. The validation of the code developed in this paper shows that the code can be adopted to do some typical transient and accident analysis. The model and code presented in this paper can be used not only in the safety analysis of pool-type sodium cooled fast reactor, but also in the development of CEFR simulation platform. (author)

  16. Cooling of core debris within the reactor vessel lower head

    International Nuclear Information System (INIS)

    Under severe accident conditions, the most crucial action for recovery from the accident state is to cool the core debris and prevent or terminate attack on the remaining fission product barriers. One means of preventing attack on the containment structures is to retain the core debris within the reactor vessel. The TMI-2 accident demonstrated that this could be accomplished by water resident within the reactor vessel combined with injection on a continual basis to quench the debris and remove decay heat over the long term. Some accident situations could result in the transport of molten core debris to the lower plenum, as occurred in TMI-2, the boiloff of water in the lower plenum, and the inability to add water to the reactor coolant system (RCS). Even in this extreme set of circumstances, sufficient cooling may be available to prevent failure of the reactor pressure vessel (RPV) lower head and thereby retain the core debris within the vessel. Containment configurations like Zion would result in substantial accumulation of water around the lower parts of the reactor vessel for most accident sequences. For some PWR containments, there could be substantial water accumulation around the reactor vessel and the hot and cold legs. If this water could directly contact the carbon steel vessel surface and RCS piping, substantial energy could be removed from the primary system and in particular the RPV lower head. Experiments discussed in this paper demonstrate nucleate boiling heat removal rates from the outer surface of a simulated RPV lower head surrounded by typical reflective insulation used in nuclear power plants

  17. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  18. Proceedings (slides) of the OECD/NEA Workshop on Innovations in Water-cooled Reactor Technologies

    International Nuclear Information System (INIS)

    development or being considered for future water-cooled reactors; - Advantages that Gen III reactors have over previous designs in terms of economics, fuel utilisation, thermal efficiency, etc; - Operational issues of nuclear power plants in future low carbon energy systems with high shares of variable renewables, and issues posed by climate change (e.g. water scarcity, increased air and water temperatures and extreme weather events). - Standardisation, modularization and constructability issues and challenges; - A discussion of key differences between Gen II and Gen III designs, and possibilities of back-fitting Generation II reactors with new technologies, as part of a Long Term Operation strategy. This document brings together the available presentations (slides), dealing with: 1 - Utility safety and performance requirements in Europe (J. Spiler); 2 - EPRI Utility Requirement Document (S.B. Kim); 3 - WENRA activities on new and existing reactors (F. Feron); 4 - Evolution of the Finnish safety regulations and implementation ( M.L. Jaervinen); 5 - Multinational Design Evaluation Programme (J. Husse); 6 - EDF France modernization program for the existing NPPs (G. Ferraro); 7 - Innovations in GEN III designs and modernisation of existing NPP - An operator's point of view (F. Bertels); 8 - Modernisation of existing NPPs in Switzerland (W. Denk); 9 - Nuclear Technology Improvements in Modernization, Refurbishment and New Build Projects in Finland (H. Tuomisto); 10 - Round table discussion - Renewable and nuclear energy-based mitigation of climate change: substitution for fossil fuel usage (M. Golay); 11 - Innovation in water cooled reactor technologies (B. de l'Epinois); 12 - APR1400 - Safe, Reliable Technology (S. W. Kim); 13 - Advanced safety features of 3. generation VVER Plants (J. Laaksonen); 14 - Additional information on modern VVER GEN III Technology (M. Maltsev); 15 - Research and Development on Advanced PWR Design Improvement and Innovation in NPI (C. Yu); 16 - GE

  19. Air conditioning system with supplemental ice storing and cooling capacity

    Science.gov (United States)

    Weng, Kuo-Lianq; Weng, Kuo-Liang

    1998-01-01

    The present air conditioning system with ice storing and cooling capacity can generate and store ice in its pipe assembly or in an ice storage tank particularly equipped for the system, depending on the type of the air conditioning system. The system is characterized in particular in that ice can be produced and stored in the air conditioning system whereby the time of supplying cooled air can be effectively extended with the merit that the operation cycle of the on and off of the compressor can be prolonged, extending the operation lifespan of the compressor in one aspect. In another aspect, ice production and storage in great amount can be performed in an off-peak period of the electrical power consumption and the stored ice can be utilized in the peak period of the power consumption so as to provide supplemental cooling capacity for the compressor of the air conditioning system whereby the shift of peak and off-peak power consumption can be effected with ease. The present air conditioning system can lower the installation expense for an ice-storing air conditioning system and can also be applied to an old conventional air conditioning system.

  20. Dry storage systems with free convection air cooling

    International Nuclear Information System (INIS)

    Several design principles to remove heat from the spent fuel by free air convection are illustrated and described. The key safety considerations were felt to be: loss of coolant is impossible as the passive system uses air as a coolant; overheating is precluded because as the temperatures of the containers rises the coolant flow rate increases; mass of the storage building provides a large heat sink and therefore a rapid temperature rise is impossible; and lack of any active external support requirements makes the cooling process less likely to equipment or operator failures. An example of this type of storage already exists. The German HTGR is operated with spherical graphite fuel elements which are stored in canister and in storage cells. The concept is a double cooling system with free convection inside the cells and heat exchange via two side walls of the cell to the ambient air in the cooling ducts. Technical description of the TN 1300 cask is also presented

  1. Liquid metal cooled reactors: Experience in design and operation

    International Nuclear Information System (INIS)

    on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

  2. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  3. Thermal computations for electronics conductive, radiative, and convective air cooling

    CERN Document Server

    Ellison, Gordon

    2010-01-01

    IntroductionPrimary mechanisms of heat flowConductionApplication example: Silicon chip resistance calculationConvectionApplication example: Chassis panel cooled by natural convectionRadiationApplication example: Chassis panel cooled only by radiation 7Illustrative example: Simple thermal network model for a heat sinked power transistorIllustrative example: Thermal network circuit for a printed circuit boardCompact component modelsIllustrative example: Pressure and thermal circuits for a forced air cooled enclosureIllustrative example: A single chip package on a printed circuit board-the proble

  4. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  5. Physical and technical aspects of lead cooled fast reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2001-07-01

    The safety analysis of lead-cooled fast reactors has been performed for the well-developed concept of BREST-OD-300 reactor. The most severe accidents have been considered. An ultimate design-basis accident has been defined as an event resulting from an external impact and involving a loss of leak-tightness of the lead circuit, loss of forced circulation of lead and loss of heat sink to the secondary circuit, failure of controls and of reactor scram with resultant insertion of total reactivity margin, etc. It was assumed in accident analysis that the protective feature available for accident mitigation was only reactivity feedback on the changes in the temperatures of the reactor core elements and coolant flow rate, and in some cases also actuation of passive protections of threshold action in response to low flow rate and high coolant temperature at the core outlet. It should be noted that the majority of the analyzed accidents could be overcame even without initiation of the above protections. It has been demonstrated that a combination of inherent properties of lead coolant, nitride fuel, physical and design features of fast reactors will ensure natural safety of BREST and are instrumental for avoiding by a deterministic approach the accidents associated with a significant release of radioactivity and requiring evacuation of people in any credible initiating event and a combination of events. (author)

  6. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  7. Convective cooling in a pool-type research reactor

    International Nuclear Information System (INIS)

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm−3. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s−1 from the 4” pipe, and predicted pool surface temperature not exceeding 30°C

  8. Heat exchanger performance in main cooling system on high temperature test operation at high temperature gas-cooled reactor 'HTTR'

    International Nuclear Information System (INIS)

    High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at Japan Atomic Energy Research Institute achieved the reactor outlet coolant temperature of 950degC for the first time in the world at Apr.19, 2004. To remove generated heat at reactor core and to hold reactor inlet coolant temperature as specified temperature, heat exchangers in HTTR main cooling system should have designed heat exchange performance. In this report, heat exchanger performance is evaluated based on measurement data in high temperature test operation. And it is confirmed the adequacy of heat exchanger designing method by comparison of evaluated value with designed value. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

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

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

    International Nuclear Information System (INIS)

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

  11. The generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    The gas cooled fast reactor (GFR) is a helium-cooled fast spectrum reactor operating within a closed fuel cycle. It combines the advantages of fast reactors, in terms of a more sustainable use of uranium resources and waste minimisation, with the wider applicability of high temperature gas reactors, in terms of high efficiency electricity generation and the co-generation of high-quality process heat. Other advantages like the absence of threshold effect due to phase changing, the optical transparency and chemical inertness of the Helium coolant are also acknowledged. Within the European Union, GFR is one of the three fast reactors proposed for development to the demonstration stage within the European Sustainable Nuclear Industry Initiative (ESNII). On a wider global scale, GFR is one of the six systems proposed for further development within the Generation IV International Forum (GIF). In this respect, France, Switzerland, Japan and the European Union (through EURATOM) are signatories to the 'System Arrangement', the instrument through which the international research efforts are coordinated. This paper presents the current status of the development of the GFR system. The status of the GFR programme in each of the signatory countries is summarised including the intended contribution of the newly launched EURATOM 7. Framework Programme project - GoFastR. France has provided the bulk of the effort on conceptual design, safety assessment and fuel development. Switzerland makes significant contributions to the GFR system in the areas of core physics, uncertainty analysis, deterministic safety assessment and fuel development. Historically Japan has been very active in the development of the GFR system. Within the Generation IV GFR system, Japan contributes to the development of fuel and core materials

  12. Study on a decay heat removal system of light water reactors using air coolers

    International Nuclear Information System (INIS)

    In the present work, a passive decay heat removal system for light water reactors (LWRs) based on a new concept is studied referring to an air cooling system (ACS) of the fast breeder reactor Monju. The present study will contribute to the reduction of severe accident risks of nuclear power plants. In this system, a blower for an air cooler (AC) is operated using the rotation of a small steam turbine by generated steam in order to cool heat transfer tubes by forced convection of air. The purpose of the present work is to investigate the plant transient caused by a station blackout (SBO) using the plant system code NETFLOW++ and decay heat removal characteristics. A calculation model is the Advanced Boiling Water Reactor (ABWR) in Japan. (author)

  13. CFD modeling of the HTR reactor cavity cooling system

    International Nuclear Information System (INIS)

    This report describes the heat transport under accident conditions from the reactor vessel wall of the INCOGEN installation to the environment. For this purpose, the heat transfer mechanisms as well as the flow patterns inside the cavity and the Reactor Cavity Cooling System (RCCS) have been calculated by the CFD (Computational Fluid Dynamics) code called CFDS-FLOW3D. The main purpose of the calculations is to determine the vessel wall temperature at which the power produced in the vessel is removed. An important assumption of the calculations is that a total of 1 MW of decay power and fission power has to be removed by the RCCS under accident conditions. In the reference calculation, about 80% of the heat is transported by radiation to the RCCS, while the remaining 20% is transported by convection of the gas in the cavity. The maximum calculated temperature on the outside of the vessel in 634 K. The reference calculation is assessed by a number of sensitivity calculations. In these calculations, the influence of the following parameters on the reactor vessel wall temperature has been determined: The turbulence model, the properties of the inlet and the outlet structures, the heat loss from the reactor vessel wall, the emissivity of structures, and the interaction between gas and radiation. Most of the parameters investigated have a small influence on the reactor vessel wall temperature. The following changes result in an increase of the reactor vessel wall temperature by 25 K or more: An increase of the heat loss from 1 MW to 2 MW, an increase of the inlet temperature from 300 K to 350 K, a decrease of the emissivity of the reactor vessel wall from 0.8 to 0.6, or very high concentrations of scattering aerosol particles. (orig.)

  14. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  15. Delayed gamma power measurement for sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Graphical abstract: Display Omitted Research highlights: →20F and 23Ne tagging agents are produced by fast neutron flux. →20F signal has been measured at the SFR Phenix prototype. → A random error of only 3% for an integration time of 2 s could be achieved. →20F and 23Ne power measurement has a reduced temperature influence. → Burn-up impact could be limited by simultaneous 20F and 23Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,α) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  16. Specific power of liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs

  17. Generation IV nuclear energy system initiative; air-cooled option RCCS studies and NSTF preparation

    International Nuclear Information System (INIS)

    The work documented in this report is a follow-on to the FY 2010 work reported on the air-cooled RCCS design of the General Atomics (GA) Modular High Temperature Gas-cooled Reactor (MHTGR). This workscope focused on confirming the scaling laws selected to design the scaled experiments in the large-scale Argonne National Laboratory (ANL) Natural Convection Shutdown Heat Removal Test Facility (NSTF) and on sensitivity studies to evaluate the effect of the modeling approximations made in FY 2010 on the models developed for the experiment design supporting analysis. In addition design support analyses were performed to evaluate NSTF configuration modifications proposed to enhance instrumentation performance. The objectives were to: (a) update and confirm the scaling relations to be used in modifying the existing NSTF facility into a scaled model of the GA-MHTGR air-cooled RCCS; and (b) provide an analytical basis for the NSTF modifications needed to conduct scaled experimental simulations of this RCCS. In summary, the scaling evaluation updated the basis that the air-cooled RCCS can be simulated at the ANL NSTF facility at a prototypic scale in the lateral direction and about half scale in the vertical direction. The density behavior of air was modified. A parallel track approach was taken with both RELAP5 system evaluations and STAR-CCM+ computational fluid dynamic (CFD) evaluations carried out as the set of instrumentation configurations were detailed.

  18. Cool Colored Roofs to Save Energy and Improve Air Quality

    Energy Technology Data Exchange (ETDEWEB)

    Akbari, Hashem; Levinson, Ronnen; Miller, William; Berdahl, Paul

    2005-08-23

    Urban areas tend to have higher air temperatures than their rural surroundings as a result of gradual surface modifications that include replacing the natural vegetation with buildings and roads. The term ''Urban Heat Island'' describes this phenomenon. The surfaces of buildings and pavements absorb solar radiation and become extremely hot, which in turn warm the surrounding air. Cities that have been ''paved over'' do not receive the benefit of the natural cooling effect of vegetation. As the air temperature rises, so does the demand for air-conditioning (a/c). This leads to higher emissions from power plants, as well as increased smog formation as a result of warmer temperatures. In the United States, we have found that this increase in air temperature is responsible for 5-10% of urban peak electric demand for a/c use, and as much as 20% of population-weighted smog concentrations in urban areas. Simple ways to cool the cities are the use of reflective surfaces (rooftops and pavements) and planting of urban vegetation. On a large scale, the evapotranspiration from vegetation and increased reflection of incoming solar radiation by reflective surfaces will cool a community a few degrees in the summer. As an example, computer simulations for Los Angeles, CA show that resurfacing about two-third of the pavements and rooftops with reflective surfaces and planting three trees per house can cool down LA by an average of 2-3K. This reduction in air temperature will reduce urban smog exposure in the LA basin by roughly the same amount as removing the basin entire onroad vehicle exhaust. Heat island mitigation is an effective air pollution control strategy, more than paying for itself in cooling energy cost savings. We estimate that the cooling energy savings in U.S. from cool surfaces and shade trees, when fully implemented, is about $5 billion per year (about $100 per air-conditioned house).

  19. Accident analysis of heavy water cooled thorium breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  20. Liquid metal cooled reactors and fuel cycles for international security

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), the University of California, Berkeley (UCB) and the Central Research Institute of Electric Power Industry (CRIEPI) have conducted several studies on small reactors. These efforts include studies on the Secure, Transportable, Autonomous Reactor (STAR), the Super, Safe, Small and Simple (4S) reactor, the Encapsulated Nuclear Heat Source (ENHS), and the ANL STAR-LM and STAR-H2 studies. Based on these studies, a concept called the Highly Secure Nuclear Fuel Cycle (HSNFC) is being introduced to focus on the fuel cycle and institutional elements of the STAR concept. The HSNFC foresees the development of small nuclear reactors suitable for remote regions and countries with limited or developing energy infrastructures. The HSNFC is intended to reduce the complexity and expense of eliminating concerns about nuclear proliferation and severe nuclear accidents, even when the user is in the initial phase of developing an energy infrastructure. It is the objective of HSNFC to make the environmental, energy security and reliability benefits of nuclear energy available to all at a competitive cost. A fundamental feature of the HSNFC and the STAR concept is that the nuclear fuel is contained within the sealed reactor vessel when shipped to the user; the spent fuel is then returned to the supplier without being removed from the reactor vessel. Small reactors with very long life nuclear cores are key to the HSNFC concept being competitive with alternative sources of energy. Evaluations completed to date have confirmed that a small liquid metal cooled reactor can likely achieve the performance necessary to make the HSNFC competitive. The ENHS and STAR-LM are examples of reactors that could be used in HSNFC and 4S, with some modification, may also be suitable. The application of these reactors in the HSNFC is discussed in this paper. The institutional conditions necessary to implement the HSNFC

  1. Accident analysis of heavy water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  2. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value

  3. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    International Nuclear Information System (INIS)

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO2 as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO2 is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year

  4. Testing Of Secondary Cooling Component Of TRIGA Mark Reactor

    International Nuclear Information System (INIS)

    The aim of this activity is to improve the knowledge of the mechanical testing technology of the research reactor cooling pipe material. The way which was chosen is through a series of testing to know the mechanical properties of carbon steel pipe used in TRIGA-MARK II secondary cooling pipe. Scopes of these testing activities are tensile testing, hardness testing, chemical composition analysis, and metallography analysis. Visual examination shows that thickness of the pipe was reduced over the range 0.31-1.76 mm and there was scales inside the pipe about 7.1-9.1 mm. Result of the mechanical testing shows that ultimate tensile strength, yield strength, elongation and. hardness of that material are 39 kg mm2, 34 kg/mm2, 38 %, and HV161, respectively. That yield strength value is on the design range

  5. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    The paper presents the current status of the Gas cooled Fast Reactor system development which is shared within the Generation IV International Forum including EURATOM through the 7th Framework Programme project GoFastR. The various areas considered will include suitable fuel compounds and high temperature resistant cladding materials options, core design optimisation, primary system boundary, energy conversion. The safety approach, mainly oriented on core cooling for the moment, will be recalled together with a discussion of the results obtained. Further potential improvements or simplification of the system safety, at the light of the Fukushima accident, including an indirect coupled cycle for the energy conversion and a self sustainable Decay Heat Removal loop will be mentioned. The main issues related to the necessary R&D programme accompanying the system development will be recalled (fuel and materials, helium coolant technology, components such as gas circulators, valves and heat exchangers, thermal barriers). (author)

  6. Shape optimization of a sodium cooled fast reactor

    Science.gov (United States)

    Schmitt, Damien; Allaire, Grégoire; Pantz, Olivier; Pozin, Nicolas

    2014-06-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth.(1, 2) Usual optimization methods for core conception are based on a parametric description of a given core design(3).(4) New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints.(5, 6) First studies show that these methods could be applied to sodium cooled core conception.(7) In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a get realistic core layout. Its caracteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas.

  7. Coolant chemistry of the advanced carbon dioxide cooled reactor

    International Nuclear Information System (INIS)

    The large scale production of electricity by uranium fission has been achieved in the United Kingdom exclusively by reactors which are gas cooled and moderated by graphite. In this way the use of uranium with close to the natural isotopic content was possible. Once the choice of graphite as moderator had been made then the selection of a suitable gas to transport heat from the core to the steam generating equipment was limited and, in fact, only two have been identified as having suitable chemical and nuclear properties, namely helium and carbon dioxide. The first of these has the disadvantage of being expensive but has a high heat transfer capability and is fundamentally inert, its reactivity being controlled entirely by the level of impurities such as hydrogen and water. With the closure of the OECD Dragon High Temperature Reactor Project interest in helium cooling for nuclear plant has faded in the UK. The alternative coolant, carbon dioxide, which is cheap but chemically reactive is used in the first generation Magnox power stations and in the Commercial Advanced Gas Cooled Reactor (CAGR) design. In the more highly rated CAGR design the fuel consists of enriched uranium in the form of dioxide encased in stainless steel; the gas outlet temperature of the core is increased to around 6000C. The purpose of this paper is to provide an insight into the current thinking about the nature of the complex chemistry associated with the CAGR coolant and how this chemistry influences the rate of deposition onto the fuel pin surfaces and the rate of graphite moderator oxidation. The application of these ideas to the prediction of the behaviour of a CAGR core with particular reference to the calculation of coolant composition within the porous moderator structure at points remote from the surface, is outlined and the use of all this information to define a satisfactory range of coolant composition is also described

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

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-10-01

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

  11. AIR COOLING IN AUTOMOBILES USING VORTEX TUBE REFRIGERATION SYSTEM

    Directory of Open Access Journals (Sweden)

    B.SREENIVASA KUMAR REDDY

    2013-02-01

    Full Text Available Refrigeration plays an important role in developing countries, primarily for the preservation of food, medicine, and for air conditioning. Conventional refrigeration systems are using Freon as refrigerant. As they are the main cause for depletion of ozone layer, extensive research work is going on alternate refrigeration systems. Vortex tube is a non conventional cooling device, having no moving parts which will produce cold air and hot air from the source of compressed air without affecting the environment. When a high pressure air is tangentially injected into vortex chamber a strong vortex flow will be created which will be split into two air streams, one hot stream and the other is cold stream at its ends.

  12. Bruce NGS A/B assessment of reactor vault fans on air mixing patterns

    International Nuclear Information System (INIS)

    The development and results of numerical simulations of air mixing patterns in the CANDU Bruce Nuclear Generating Station reactor vault, as a function of vault cooling fan combinations, are presented. The results of this analysis will serve as a basis for selecting appropriate fan combination scenarios to consider in the upcoming post-LOCA (loss of coolant) hydrogen-air-steam mixing analysis. Following a severe reactor accident in which fuel cooling is impaired, a significant amount of hydrogen may be produced from the steam/Zircaloy reaction and subsequently released into containment. The hydrogen ignition system mitigates the consequences of hydrogen burns to within acceptable safety limits. Igniters deliberately initiate a burn of the hydrogen-air-steam mixture as it reaches its flammability limits. However without adequate mixing, the igniters may become blinded by a region of non-flammable hydrogen mixture while an unfavourable hydrogen mixture forms elsewhere. The vault cooling fans play an important role in promoting mixing in the vault atmosphere. To help assess the effects of vault cooling fans on air mixing, an analysis was carried out to identify the air mixing patterns as a function of different fan availability combinations. The three-dimensional containment code, GOTHIC, was used to model the Bruce containment with modelling emphasis on the reactor vault geometry and the vault cooling system fans. Twenty-five fan combination air mixing simulations and eight tracer gas fan dispersion simulations were performed. The results showed that air mixing patterns created by individual fans can be superimposed to determine the effects of various fan combinations, there was symmetry of flow patterns between the west and east vault halves, and there was a general absence of significant stagnant regions in the reactor vault. 8 refs., 7 figs., 3 tabs

  13. Seismic study on high temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    The resistance against earthquakes of a high temperature gas-cooled reactor (HTGR) core with block-type fuel is not yet fully ascertained. Seismic studies must be made if such a reactor plant is to be installed in the areas with frequent earthquakes. The experimental and analytical studies for the seismic response of the HTGR core were carried out. First, the fundamental behavior, such as the softening characteristic of a single stacked column (which is piled up with blocks) and the hardening characteristic with the block impact were clarified from the seismic experiments. Second, the displacement and the impact characteristics of the two-dimensional vertical core and the two-dimensional horizontal core were studied from the seismic experiments. Finally, analytical methods and computer programs for the seismic response of HTGR cores were developed. (author) 57 refs

  14. Cooling molten salt reactors using “gas-lift”

    Energy Technology Data Exchange (ETDEWEB)

    Zitek, Pavel, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Valenta, Vaclav, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Klimko, Marek, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz [University of West Bohemia in Pilsen, Univerzitní 8, 306 14 Pilsen (Czech Republic)

    2014-08-06

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a “Two-phase flow demonstrator” (TFD) used for experimental study of the “gas-lift” system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for “gas-lift” (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  15. Cooling molten salt reactors using "gas-lift"

    Science.gov (United States)

    Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

    2014-08-01

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

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

    International Nuclear Information System (INIS)

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

  17. Cooling molten salt reactors using “gas-lift”

    International Nuclear Information System (INIS)

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a “Two-phase flow demonstrator” (TFD) used for experimental study of the “gas-lift” system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for “gas-lift” (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors

  18. Thermal top shield for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Proposed is a thermal top shield for gas-cooled nuclear reactors which together with the thermal side and bottom shield forms an almost gas-tight room for taking up the core structure and which protects the top of the concrete vessel sufficiently against overheating. The thermal top shield consists of top shield elements put closely together, which are made of at least two horizontal metal layers and at least one moderator layer located between the metal layers and which are fixed to the top liner by means of drawbars. (orig.)

  19. Liquid metal cooled fast breeder nuclear reactor constructions

    International Nuclear Information System (INIS)

    A description is given of a liquid metal cooled fast breeder nuclear reactor construction of the pool kind in which the primary vessel incorporates an annular yoke fabricated from arcuate segments. The yoke is suspended from the roof structure of the vault by a first annular series of tie straps arranged outside the primary vessel whilst a strongback on which the fuel assembly sits inside the primary vessel is supported from the yoke by a second series of tie straps. The yoke has upwardly and downwardly extending legs which are extended by upper and lower strakes respectively of the primary vessel. (U.K.)

  20. Materials challenges for the Supercritical Water-Cooled Reactor (SCWR)

    International Nuclear Information System (INIS)

    This paper discusses the materials requirements of the Supercritical Water-cooled Reactor (SCWR) which arise from its severe expected operating conditions: (i) Outlet Temperature (to 650 C); (ii) Pressure of 25 MPa for the coolant containment, (iii) Thermochemical stress in the presence of supercritical water, and (iv) Radiative damage (up to 150 dpa for the fast spectrum variant). These operating conditions are reviewed; the phenomenology of materials in the supercritical water environment that create the materials challenges is discussed; knowledge gaps are identified, and efforts to understand material behaviour under the operating conditions expected in the SCWR are described. (author)

  1. Gas-cooled reactor coolant circulator and blower technology

    International Nuclear Information System (INIS)

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

  2. Thorium fueled high temperature gas cooled reactors. An assessment

    International Nuclear Information System (INIS)

    The use of thorium as a fertile fuel for the High Temperature Gas Cooled Reactor (HTR) instead of uranium has been reviewed. It has been concluded that the use of thorium might be beneficial to reduce the actinide waste production. To achieve a real advancement, the uranium of the spent fuel has to be recycled and the requested make-up fissile material for the fresh fuel has to be used in the form of highly-enriched uranium. A self-sustaining fuel cycle may be possible in the HTR of large core size, but this could reduce the inherent safety features of the design. (orig.)

  3. Passive subsystem of emergency core cooling of pressurized water reactor

    International Nuclear Information System (INIS)

    Between the accident accumulator, resp. the storage tank and the primary circuit or the reactor an injector is inserted in the pipe of cooling borated water whose propelling nozzle is directly or indirectly connected to the secondary side of the steam generator, resp. to the secondary circuit of the power plant. In the steam supply pipe between the steam generator and the accident accumulator is located a pressure reducing supply valve. In the pipe of the borated water a heat exchanger is placed before the injector. (M.D.)

  4. Study on forced air convection cooling for electronic assemblies

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    The slotted fin concept was employed to improve the air cooling performance of plate-fin in heat sinks.Numerical simulations of laminar heat transfer and flow pressure drop were conducted for the integral plate fin,discrete plate fin and discrete slotted fin heat sinks.It is found that the performance of the discrete plate fin is better than that of the integral continuum plate fin and the performance of slotted fin is better than that of the discrete plate fin at the same pumping power of the fan.A new type of heat sink characterized by discrete and slotted fin surfaces with thinner fins and smaller spaces between fins is then proposed.Preliminary computation shows that this type of heat sink may be useful for the next generation of higher thermal load CPUs.The limit of cooling capacity for air-cooling techniques was also addressed.

  5. SIMMER-III modeling of gas cooled fast reactor

    International Nuclear Information System (INIS)

    This paper deals with extension and application of the SIMMER-III code for safety studies of a gas cooled fast reactor. The equation of state of the helium gas and its thermal physical properties have been prepared and implemented in the code. The geometric, thermal hydraulic and neutronic models have been set up for the ALLEGERO reactor. The code and the associated model are verified by comparing steady state and unprotected loss of flow 20% remained flow rate (ULOF-20%) results with those done by other project partners. Reasonable or good agreements have been achieved for major physical variables. The unprotected loss of coolant accident (ULOCA) case is a severe transient case with core melting and degradation that was emulated only by SIMMER, in the project. In the initiating phase the clad becomes molten, this triggers the first power excursion. Then the fuel becomes more mobile and further power excursions take place, which lead to core melting and degradation. The fuel is ejected by power excursion and then moves relatively slowly to the lower part of vessel. Finally there are only a few kilograms of fuel escaping to the vessel outside (into reactor container) and the released thermal energy is about 6 GJ within a period of one minute. The final power stays below one MW and the reactor is in a deep sub-criticality state, since 1/2 fuel becomes noneffective. (author)

  6. Air cooling effect of fins on a Honda shine bike

    Directory of Open Access Journals (Sweden)

    Padhiyar Abhesinh J

    2015-05-01

    Full Text Available The main of aim of this work is to study various researches done in past to improve heat transfer rate of cooling fins by changing cylinder block fin geometry. Low rate of heat transfer through cooling fins is the main problem in this type of cooling. So efficiency of the engine is increase by increase the heat transfer. Examples of direct air cooling in modern automobiles are rare. The most common example is the commercials Automobile bike like a Honda Shine, Bajaj bike, Honda splendor etc. It is conclude about shape try to this fins is more effectively heat transfer in Honda shine bike compare to existing fins. After FEA Analysis it checking on fin whether efficiency of heat transfer increases or not. This work validation with Experimental and Mathematical.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-14

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

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

    International Nuclear Information System (INIS)

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

  9. 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.

  10. Radioactivity evaluation code system for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    A code system for the evaluation of the behavior of radioactive fission products (FP) in high temperature gas-cooled reactors (HTGR) is described. The first half of this report is devoted to the description of the conceivable behavior of FPs in the experimental very high temperature gas-cooled reactor being designed at JAERI. The transfer of FPs from the fuel to the primary coolant is considered in three steps; the release of FPs from the coated fuel particles; the diffusion of FPs within graphite sleeves; and the transfer of FPs from the sleeve surface to the coolant. As for the FP behavior within the primary coolant system, the deposition of FPs on various walls of the system is considered. As for the secondary and the thermal utilization systems, the transfer of tritium is specially considered. The calculation model for the transfer and deposition of fission products within the whole plant system is presented by a chart. The second half of this report describes the evaluation code system. The physical and mathematical models treated in each component code are presented and discussed. (Aoki, K.)

  11. ELECTRA: A European Lead-bismuth Cooled Training Reactor

    International Nuclear Information System (INIS)

    The lack of low power liquid metal cooled reactors has meant that few engineers within the nuclear power industry and research community are familiar with operational procedures of this family of coolants, expected to be used for Generation IV fast neutron systems. The reasons for this lack may include safety issues related to use of Mercury, NaK or sodium as applied in early low power reactors. Especially in western Europe, no low power liquid metal cooled reactor was ever in operation. Here, we present the design of a 2 MWth lead-bismuth cooled reactor with (Pu,Zr)N fuel, relying on natural convection for full power operation. The combination of low power density with natural convection for heat removal makes the reactor ideal for training purposes. The large thermal expansion of heavy liquid metals makes is possible to design low power fast neutron reactors relying on natural convection. Since at present, there exists no suitable material for pumps operating at high velocity in lead alloy environments, the only possible short term solution for constructing such a reactor is anyway to design for 100% natural circulation of the coolant. For this purpose, a small core height, a large difference between coolant inlet and outlet temperatures and a low coolant velocity is desired. By application of (Pu,Zr)N fuel, criticality can be achieved with a fissile inventory of 100 kg LWR grade plutonium. Monte Carlo simulations show that 19 hexagonal fuel assemblies, each with 91 fuel pins having an outer diameter of 1.1 cm, and an active height of 15 cm is sufficient to obtain a critical core. Including end pellets, gas plenum and end caps, the total core height is limited to 30 cm. Adopting P/D = 1.25 and a heat exchanger elevation of 4 m, it is found that 2 MW of thermal power may be removed by a natural circulation velocity of 0.4 m/s. This corresponds to a linear rating of 8 kW/m and a temperature increase of the coolant equal to 240 degrees. Limiting the clad temperature

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

    International Nuclear Information System (INIS)

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

  13. Mitigation of gas entrainment in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Argon cover gas may entrain into sodium in the hot pool and in the surge tank of liquid Sodium cooled Fast Reactor (SFR) due to various mechanisms. The entrained cover gas may hinder the normal reactor operation in many ways such as reduction in heat transfer in the heat exchanger, causing neutronic perturbation inside the core etc. The recirculating gas bubbles also enhance chances of cavitation in the pumps. Therefore, it is required to mitigate gas entrainment in reactor. High free surface velocity and turbulence level are the causes of gas entrainment from free surface. Hence, gas entrainment can be avoided by reducing these factors at free surface. This is achieved by employing gas entrainment mitigation devices which essentially alters the flow pattern and reduces velocity and turbulence level at the free surface. A combined experimental and computational approach is proposed to develop such devices. The computational model which is used for parametric studies is first validated against experiments carried out in scale down water model of reactor primary circuit and surge tank. The effect of different gas entrainment mitigating devices for reduction in free surface velocity and turbulence level has been analyzed using this validated CFD model. Based on the CFD analyses, the final geometry of the gas entrainment mitigating devices has been selected and optimized. Finally the selected devices have been tested to confirm its performance for mitigation of gas entrainment in large scale models of reactor primary circuit and surge tank. This paper presents the studies carried out towards development of suitable gas entrainment mitigation devices for hot pool and surge tank of SFR. (author)

  14. Passive Safety Optimization in Liquid Sodium-Cooled Reactors

    International Nuclear Information System (INIS)

    The governing equations and the solutions schemes are developed for the three-dimensional thermal-hydraulic model.A detailed constitutive relations are also developed through the analysis with a CFD code. The developed model is able to obtain a detailed thermal hydraulic information in a subassemlby of a liquid metal-cooled reactor core. The model has been integrated with the system analysis codes SASSYS-1 and SSC-K to be validated for the SHRT-17 test performed in the EBR-II reactor. The baseline analyses were performed with the customary thermal-hydraulic model and with the new model for the reference design of KALIMER-150. The results point out that increased detail in the thermal-hydraulic analysis must be matched by increased detail in the reactivity feedback modeling, especially the radial core expansion model, in order to provide distinguishable differences. The analyses performed to identify certain key safety features have indicated that the adjustment of core restraint system is able to provide beneficial bending of hexcans, resulting in enhanced negative reactivity feedback in unprotected accidents. Tasks were performed to identify and assess the implications for plant safety of proposed specific approaches for reducing capital and operating costs of next generation sodium-cooled fast reactors. A new plant design of the sodium-cooled reactor concept is developed utilizing a gas turbine Brayton cycle, which uses supercritical carbon dioxide (S-CO2) as the working fluid. In addition, the design innovations are incorporated in modular sodium-to-supercritical carbon dioxide heat exchangers that enable the traditional intermediate heat transport circuit to be eliminated. Several evaluations and analyses for the safety design and efficiency suggest that the developed system is safe and cost-effective. Test plans are developed for the measurement of phenomenological data describing freezing of molten metallic fuel, melt relocation and interaction with steel

  15. Optimal air-supply mode of hybrid system with radiant cooling and dedicated outdoor air

    Institute of Scientific and Technical Information of China (English)

    丁研; 田喆; 朱能

    2015-01-01

    The hybrid system with radiant cooling and dedicated outdoor air not only possesses high energy efficiency, but also creates a healthy and comfortable indoor environment. Indoor air quality will be improved by the dedicated outdoor air system (DOAS) and indoor thermal comfort can be enhanced by the radiant cooling system (RCS). The optimal air-supply mode of the hybrid system and the corresponding design approach were investigated. A full-scale experimental chamber with various air outlets and the ceiling radiant cooling panels (CRCP) was designed and established. The performances of different air-supply modes along with CRCPs were analyzed by multi-index evaluations. Preliminary investigations were also conducted on the humidity stratification and the control effect of different airflow modes to prevent condensation on CRCP. The overhead supply air is recommended as the best combination mode for the hybrid system after comprehensive comparison of the experiment results. The optimal proportion of CRCP accounting for the total cooling capacities in accord with specific cooling loads is found, which may provide valuable reference for the design and operation of the hybrid system.

  16. BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G. [N.A. Dollezhal Institute ' NIKIET' , PO Box 788, Moscow, 101000 (Russian Federation)

    2006-07-01

    This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with {beta}{sub eff}, low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

  17. CEA program on future generation light water modular reactors and gas cooled reactors

    International Nuclear Information System (INIS)

    The CEA programme on 'Future Generation Reactors and Fuel Cycles' aims at studying and developing the mean and long term most promising options for nuclear reactors, fuels and reprocessing. These options should contribute to make the nuclear energy a major source of the sustainable development. The program also aims at maintaining at the highest level of competency the technologies with which the CEA will be able to bring to national achievements or international projects in the next decades, projects whose specifications and calendar are today unknown. These studies on the 'Future Generation Reactors and Fuel Cycles' constitute a field privileged for international collaboration. The corresponding researches are structured in four main axes: Innovations for LWR; Systems of 4th generation; Sodium-cooled reactors; Systems which are the object for survey or exploratory studies. Studies on future nuclear gas technologies are mainly covered by the 4th generation programme (Gen IV). Within this context, the goals pursued, in particular the minimization of the production of long lived waste and the saving of resources (i.e.: the optimised utilisation of fissile and fertile nuclear fuels), could justify an evolution towards hard neutron spectra and high temperatures, to cover applications other than the electricity production, e.g.: hydrogen production, desalination, cogeneration. The main R and D axis for these long-term objectives currently the area of Gas Cooled Reactors (GCR). The corresponding program is structured through eight R and D projects details of which are presented within the paper. (author)

  18. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    International Nuclear Information System (INIS)

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540 C and the helium coolant was delivered at 7 MPa at 625-925 C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the high temperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  19. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    Energy Technology Data Exchange (ETDEWEB)

    Lee O. Nelson

    2011-04-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  20. Present Status of Canadian Organic-Cooled Reactor Technology

    International Nuclear Information System (INIS)

    Canada has been operating the first organic-cooled D2O-moderated pressure-tube reactor (WR-1) since November 1965. The operation of WR-1 and the supporting development programmes at the Whiteshell Nuclear Research Establishment are of direct interest to those seeking to exploit the advantages of organic coolants. The following contributions have been made which bear on the technical feasibility of the concept. 1. HB-40, a partially hydrogenated terphenyl mixture, has been used routinely in WR-1 at temperatures up to 370°C. It is now developed to the point where it is the major contender for use in a power reactor. It has the advantages of being liquid at room temperature and offering cost savings over other organic coolants at comparable conditions. 2. Zirconium base alloys can be made compatible with organic coolants provided 50 to 300 ppm of water are added to condition the surface of the zirconium alloy against hydrogen penetration. Chlorine contamination must be controlled. 3. Fuel can be operated at practical ratings without limitations imposed by fouling provided reasonable attention is paid to coolant chemistry. 4. Non-bonded uranium carbide fuels have been taken to bumups as high 11 000 MWd/t and would probably be acceptable at burnups in excess of 16 000 MWd/tU. These and other subjects associated with the present status of the technology are the subject of this paper. The excellent performance of WR-1 indicates that, as yet, no insurmountable technical limitation faces the designers of organic cooled reactors. (author)

  1. Reactor core of a gas-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    In order to increase the outlet temperature of the coolant (helium) leaving the reactor core of a gas-cooled high-temperature nuclear reactor and thereby to improve its thermal efficiency there is proposed to design the geometry of the fuel elements or the fuel element units and/or their main dimensions non-uniformly. Those fuel elements whose geometry causes a larger pressure drop of the coolant gas are to be arranged towards the outlet side of the hot coolant. (GL) 891 GL/GL 892 MKO

  2. Incorporation of statistical distribution of particle properties in chemical reactor design and operation: the cooled tubular reactor

    OpenAIRE

    Wijngaarden, R.J.; Westerterp, K.R.

    1992-01-01

    Pellet heat and mass transfer coefficients inside packed beds do not have definite deterministic values, but are stochastic quantities with a certain distribution. Here, a method is presented to incorporate the stochastic distribution of pellet properties in reactor design and operation models. The theory presented is illustrated with a number of examples. It is shown that pellet-scale statistics have an impact on cooled tubular reactor design and operation. Cooled tubular reactor design is d...

  3. Hybrid high temperature gas-cooled reactor, thermonuclear fusion

    International Nuclear Information System (INIS)

    The project of a multi-purpose high temperature gas-cooled reactor started in 1969. The Atomic Energy Commission, Japan, approved in 1980 the budget for the design study of the experimental reactor. The conceptual design is in progress. The manufacturing of coated fuel pellets and the test method have been developed. The study of graphite structure is carried out. Corrosion and creep tests are made to obtain the knowledge concerning the metals in high temperature helium gas. The engineering study of various machines and structures operating at high temperature is performed. International cooperative works are considered. The experimental reactor will be critical in 1987. A critical plasma test facility, JT-60, has been constructed at the Japan Atomic Energy Research Institute. As the theoretical work on plasma confinement, the evaluation of the critical beta value of JT-60 was made. By high temperature neutral beam injection, the slowing down and heating processes of high energy particles are studied. The development of a non-circular cross-section tokamak is in progress. The construction of JT-60 will be completed in 1984. Study concerning superconducting magnets is considered. Japan is one of the members of INTOR project. (Kato, T.)

  4. Romanian Contribution to the Development of Lead Cooled Fast Reactors

    International Nuclear Information System (INIS)

    In Romania the nuclear energy is considered an important component of energy mix and for a country sustainable development. Presently based on PHWR CANDU technology, the research and development activities, part of the national nuclear power programme, provide an increased effort towards generation 4, dedicated to support fast reactor lead technology. In the European framework (EU R&D Framework Programmes, European Sustainable Nuclear Industry Initiative) devoted to the development of GenerationIV technologies, Romania is contributing as a partner in EU R&D projects together with a large number of EU research organizations and in the ALFRED MoU, having ANSALDO Nucleare, ENEA and INR as initial members. ALFRED (Advanced Lead Fast Reactor European Demonstrator) project aims to build a 125MWe lead cooled fast reactor demonstrator, connected to the electrical network, with a target date for operation start-up in 2025. In February 2011 Romanian Government approved the option to host ALFRED demonstrator. Based on the access to European structural funds, existing nuclear experience and EU orientation to build the demonstrators in the new member states, Romania is an important option for siting process. An investigation of the existing national capabilities, identification of additional infrastructure and identification of the professional development needs in order to prepare the siting national support are presented in the paper. The main approaches and needed resources to meet expected requirements for ALFRED implementation are discussed as well. (author)

  5. New deployment of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

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

  7. Can-rupture detection in gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author)

  8. Degradation of zirconium alloys in water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Zirconium alloys are used almost exclusively as fuel cladding, and other structural components in the cores of water-cooled nuclear reactors. This choice was largely forced on the designers by constraints to economize on neutron absorption. Among the low neutron cross-section metals (Be, Mg, Al, Si, Zr), zirconium is the only one with both adequate corrosion resistance to --3000C water, and adequate strength and ductility. However, it does suffer degradation by a number of routes; corrosion from the outside; corrosion and stress corrosion cracking from the inside; and irradiation induced embrittlement and delayed hydride cracking from within. These various modes of degradation are discussed here. Despite these possible modes of failure the in-reactor behaviour of zirconium alloy fuel cladding and other components has been generally good. However, the desire to increase fuel burnup in order to reduce fuelling costs, and a trend to higher water temperature in reactors, is pushing the zirconium alloy components to the limits of the endurance. Some discussion of future trends is given

  9. Quality assurance of emergency core cooling system in nuclear reactors

    International Nuclear Information System (INIS)

    Ensuring integrity of nuclear fuel clad is most important from radiation safety point of view. The paper provides introduction to experimental and theoretical methods for evaluation of rewetting velocity. Quality Assurance (QA) checks on Emergency Core Cooling System (ECCS) in Nuclear Reactors are very important to ensure accurate coolant flow introduction and minimum radiation hazard during Loss of Coolant Accident (LOCA). In depth knowledge of acceptable rate of temperature rise of fuel subsequent to LOCA and having fully reliable method to ensure that the same will not exceed the set limit is testimony of safe reactor operation. Spread of radioactive contamination and resultant radiation exposure from above contamination in nuclear reactors depends heavily on size and shape of split or rupture in clad. Suggests that a plant operating with 0.125 percent pin hole in fuel clad defects showed in general, upto five-fold increase in contamination level and resultant whole body radiation exposure rates in some areas of the plant when compared to a sister plant with high integrity fuel. The checks on ECCS will protect environment and public from radiation exposure to remarkable extent. (author)

  10. Air Cooling for High Temperature Power Electronics (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Waye, S.; Musselman, M.; King, C.

    2014-09-01

    Current emphasis on developing high-temperature power electronics, including wide-bandgap materials such as silicon carbide and gallium nitride, increases the opportunity for a completely air-cooled inverter at higher powers. This removes the liquid cooling system for the inverter, saving weight and volume on the liquid-to-air heat exchanger, coolant lines, pumps, and coolant, replacing them with just a fan and air supply ducting. We investigate the potential for an air-cooled heat exchanger from a component and systems-level approach to meet specific power and power density targets. A proposed baseline air-cooled heat exchanger design that does not meet those targets was optimized using a parametric computational fluid dynamics analysis, examining the effects of heat exchanger geometry and device location, fixing the device heat dissipation and maximum junction temperature. The CFD results were extrapolated to a full inverter, including casing, capacitor, bus bar, gate driver, and control board component weights and volumes. Surrogate ducting was tested to understand the pressure drop and subsequent system parasitic load. Geometries that met targets with acceptable loads on the system were down-selected for experimentation. Nine baseline configuration modules dissipated the target heat dissipation, but fell below specific power and power density targets. Six optimized configuration modules dissipated the target heat load, exceeding the specific power and power density targets. By maintaining the same 175 degrees C maximum junction temperature, an optimized heat exchanger design and higher device heat fluxes allowed a reduction in the number of modules required, increasing specific power and power density while still maintaining the inverter power.

  11. A carbon dioxide partial condensation direct cycle for advanced gas cooled fast and thermal reactors

    International Nuclear Information System (INIS)

    A carbon dioxide partial condensation direct cycle concept has been proposed for gas cooled fast and thermal reactors. The fast reactor with the concept are evaluated to be a potential alternative option to liquid metal cooled fast reactors, providing comparable cycle efficiency at the same core outlet temperature, eliminating the safety problems, simplifying the heat transport system and making easier plant maintenance. The thermal reactor with the concept is expected to be an alternative solution to current high temperature gas cooled reactors (HTGRs) with helium gas turbines, allowing comparable cycle efficiency at the moderate temperature of 650 C instead of 800 C in HTGRs. (author)

  12. Overview of PbBi-Cooled Reactor Development and ADS Program in China

    International Nuclear Information System (INIS)

    Most of China's electricity is produced from fossil fuels. Rapid growth in demand has given rise to power shortages, and the reliance on fossil fuels has led to much air pollution. Nuclear power has an important role for the China energy safety and environmental protection. According to the current layout of nuclear power development, by around 2040, pressurized water reactors are expected to level off at 200 GWe. However, the nuclear fuel shortage and the spent fuel accumulation will restrict the healthy development of nuclear energy in China. So the innovative nuclear system is expected to solve these problems. Accelerator driven sub-critical system (ADS) is an important approach to incinerate the long-lived high-level nuclear waste, which is focused by many countries. Liquid Lead-Bismuth eutectic (LBE) has many unique nuclear, thermophysical and chemical attributes that is potential candidate coolant for ADS subcritical reactors. In addition, lead and bismuth can produce copious spallation neutrons when bombarded with energetic protons. This makes LBE one of the top candidates for a high-power spallation target and coolant in an accelerator-driven subcritical system. In China, the fundamental research of ADS was carried out 10 years ago supported by Ministry of Science and Technology. From 2009, Chinese Academy of Sciences (CAS) carried out a project of the ADS prophase research to develop the key technologies about superconducting accelerator, LBE loop, materials et al. In 2011, a large scaled ADS development program to transmute the nuclear waste has been launched by CAS named ''advanced nuclear fission energy''. In this program, CAS plan to develop the ADS system by 3 phases. In the first phase, form 2011-2017, an ADS verification facility named ADS-VF will be build. The ADS-VF consists of a LBE cooled reactor coupled with a proton accelerator and liquid metal spallation target. For the second phase in this program, an ADS experimental reactor named ADS

  13. Power deposition distribution in liquid lead cooled fission reactors and effects on the reactor thermal behaviour

    International Nuclear Information System (INIS)

    In the framework of an ADS study (Accelerator Driven System, a reactor cooled by a lead bismuth alloy) the distribution of the deposited energy between the fuel, coolant and structural materials was evaluated by means of Monte Carlo calculations. The energy deposition in the coolant turned out to be about four percent of the total deposited energy. In order to study this effect, further calculations were performed on water and sodium cooled reactors. Such an analysis showed, for both coolant materials, a much lower heat deposition, about one percent. Based on such results, a thermohydraulic analysis was performed in order to verify the effect of this phenomenon on the fuel assembly temperature distribution. The main effect of a significant fraction of energy deposition in the coolant is concerned with the decrease of the fuel pellet temperature. As a consequence, taking into account this effect allows to increase the possibilities of optimization at the disposal of the designer

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

    International Nuclear Information System (INIS)

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

  15. Report of study 7.3: cooling and air conditioning

    Energy Technology Data Exchange (ETDEWEB)

    Russo, F.

    2000-07-01

    This report describes the results of the study carried out by the study group 7.3 in the triennium 1997-2000. The study was focused on industrial refrigeration and air conditioning for the large building utilising natural gas. The goal of this study, carried out in collaboration of the members of study group 7.3, was to analyse the markets of industrial refrigeration and air conditioning for large buildings to identify possibilities to increase the natural gas share in these sectors. The available technology in the two sectors of the market are described in a single section, i.e. the 'State of the art of the technology'. In this section, technical characteristics, applications, performances, new developments and others topics are discussed for absorbers, gas engines, gas turbines and fuel cells. In the 'Industrial Refrigeration' section an analysis of the present global market for the industrial sector is presented. Economics, advantages and barriers to gas units compared with the electrical units are discussed. Information on existing industrial plants, possible application options and new technology developments are described as well. The 'Air conditioning for the large building' section deals with offices, hotels, commercial buildings, hospitals and shopping centres with a cooling capacity of 350 kW or higher. It appears that the use of natural gas for cooling of large buildings has been increasing during the last decade, thanks to the greater availability of natural gas and the development of new technologies. A marketing survey of gas air-conditioning was carried out in cooperation with a group of Intergas Marketing. Based on the survey, the report describes the market position of natural gas relative to electricity. It provides the strategic prospects for further developing natural gas as a competitive option for air-conditioning of large buildings using a combination of state-of-the-art technologies. It is important to highlight

  16. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  17. Evaluating Fault Detection and Diagnostics Protocols Applied to Air-Cooled Vapor Compression Air-Conditioners

    OpenAIRE

    Yuill, David P.; Braun, James E.

    2012-01-01

    Fault detection and diagnostics (FDD) tools are being increasingly applied in air-conditioning systems. There are many different protocols used in these FDD tools, so an important question to ask is: how well do the protocols work? This paper describes the ongoing development of the first standardized method of evaluation for FDD protocols applied to air-cooled vapor compression air-conditioning systems. The general approach is to feed a library of data – including temperatures, pressures, an...

  18. the effect of design changes of cooling tower on the performance of ETRR-2 reactor

    International Nuclear Information System (INIS)

    the egyptian testing and research reactor(ETRR-2) were established in 1998 with maximum power 22 MW for research purposes. two induced draft cooling towers with different specifications was installed inh the system, a replaced (old) cooling tower was in operation from 1998 to 2003, and the present (new) cooling tower was in operation from 2003 till now. the reactor was put into operation since 1998 but it faced a lot of problems in the cooling system concerning with the thermal load dissipation. some efforts guided to study this problem to evaluate the old and present cooling tower to decide if the present cooling tower achieves a good performance in the reactor cooling system and to know why the old cooling tower have to be replaced ? and to avoid thermal problems in the future to satisfy the stable operation. in this work the study of the cooling system of the ETRR-2 is presented. this study is based on analytical, numerical and measurement investigations of the cooling system following three parts. he first part depicts the analytical solution of integrated cooling system of the reactor, the second part depicts the numerical solution of the cooling tower packing , and the third part is the evaluation of the cooling system using cooling technology, institute procedure (CTI)

  19. Thermal analysis of air-cooled fuel cells

    OpenAIRE

    Shahsavari, Setareh

    2011-01-01

    Temperature distribution in a fuel cell significantly affects the performance and efficiency of the fuel cell system. Particularly, in low temperature fuel cells, improvement of the system requires proper thermal management, which indicates the need for developing accurate thermal models. In this study, a 3D numerical thermal model is presented to analyze the heat transfer and predict the temperature distribution in air-cooled proton exchange membrane fuel cells (PEMFC). In the modeled fuel c...

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

    International Nuclear Information System (INIS)

    The Federal Target Program (FTP) of Russian Federation 'Nuclear Energy Technologies of the New Generation for 2010-2015 and for Perspective up to 2020' is aimed at development of advanced nuclear energy technologies on the basis of closed fuel cycle with fast reactors. There are advanced fast reactor technologies of the 4. generation with liquid metal cooled reactors. Development stages of maturity of fast sodium cooled reactor technology in Russia includes experimental reactors BR-5/10 (1958-2002) and BOR-60 (since 1969), nuclear power plants (NPPs) with BN-350 (1972-1999), BN-600 (since 1980), BN-800 (under construction), BN-1200 (under development). Further stage of development of fast sodium cooled reactor technology in Russia is commercialization. Lead-bismuth eutectic fast reactor technology has been proven at industrial scale for nuclear submarines in former Soviet Union. Lead based technology is currently under development and need for experimental justification. Current status and prospects of State Corporation 'Rosatom' participation in GIF activities was clarified at the 31. Meeting of Policy Group of the International Forum 'Generation-IV', Moscow, May 12-13, 2011. In June, 2010, 'Rosatom' joined the Sodium Fast Reactor Arrangement as an authorized representative of the Russian Government. It was also announced the intention of 'Rosatom' to sign the Memorandum on Lead Fast Reactor based on Russia's experience with lead-bismuth and lead cooled fast reactors. In accordance with the above FTP some innovative liquid metal cooled reactors of different design are under development in Russia. Gidropress, well known as WER designer, develops innovative lead-bismuth eutectic cooled reactor SVBR-100. NIKIET develops innovative lead cooled reactor BRESTOD-300. Some other nuclear scientific centres are also involved in this activity, e.g. Research and Development Institute for Power Engineering (RDIPE). Optimum

  1. Effect mechanism of air deflectors on the cooling performance of dry cooling tower with vertical delta radiators under crosswind

    International Nuclear Information System (INIS)

    Highlights: • A 3D numerical model was set for NDDCTV to study the effect of air deflectors. • The air deflectors improve the tower performance by 1.375 °C at uc = 6 m/s for a case. • The air deflectors reduce the air inflow deviation angle θd at most delta entries. • The reduced θd can improve the cooling performance of former deteriorated columns. • Both the radial inflow air velocity and θd impact the cooling performance of delta. - Abstract: To study the effect mechanism of air deflectors on dry cooling tower, a three dimensional numerical model was established, with full consideration of the delta structure. The accuracy and credibility of dry cooling tower numerical model were validated. By numerical model, the average air static pressure and the average radial inflow air velocity were computed and analyzed at delta air entry, sector air entry and exit faces. By the air inflow deviation angle θd, the effect of air deflectors on the aerodynamic field around tower was analyzed. The water exit temperatures of θ−1 columns, θ+2 columns and cooling sectors were also presented to clarify the effect of air deflectors. It was found that the air deflectors improved the aerodynamic field around cooling columns. The reduced air inflow deviation degree at delta entry improved the cooling performance of deteriorated columns. Referring to the radial inflow air velocity ura and the air inflow deviation degree at delta entry, the effect mechanism of air deflectors are clarified under crosswind

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

    International Nuclear Information System (INIS)

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

  3. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

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

  4. Phytoplankton distribution in three thermally distinct reactor cooling reservoirs

    International Nuclear Information System (INIS)

    Phytoplankton community structure was studied in relation to physicochemical characteristics of three South Carolina reservoirs in close proximity and of similar age and bottom type. Thermal alteration, resulting from the input of cooling water from a nuclear reactor, was substantially different in each reservoir. This provided an opportunity to compare water temperature effects separated from season. Water temperature (when examined independently in statistical models) appeared to be less important than other environmental variables in determining phytoplankton community structure. Pond C, a reservoir receiving intensely heated effluent (> 200C ΔT), displayed low species diversity (Shannon-Weaver H 0C in summer. Par Pond, having a maximum ΔT of 50C, displayed no temperature-induced alteration of phytoplankton community structure

  5. Auxiliary bearing design considerations for gas cooled reactors

    International Nuclear Information System (INIS)

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

  6. Numerical simulation of sodium pool fires in liquid metal-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    In Liquid Metal-Cooled Fast Breeder Reactor (LMFBR), the leakage of sodium can result in sodium fires. Due to sodium's high chemical reactivity in contact with air and water, sodium fires will lead to an immediate increase of the air temperature and pressure in the containment. This will harm the integrity of the containment. In order to estimate and foresee the sequence of this accident, or to prevent the accident and alleviate the influence of the accident, it is necessary to develop programs to analyze such sodium fire accidents. Based on the work of predecessors, flame sheet model is produced and used to analyze sodium pool fire accidents. Combustion model and heat transfer model are included and expatiated. And the comparison between the analytical and experimental results shows the program is creditable and reasonable. This program is more realistic to simulate the sodium pool fire accidents and can be used for nuclear safety judgement. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-09-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-09-01

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

  10. EVALUATION OF COOLING INSTRUMENTATION SYSTEM OF TRIGA MARK II REACTOR OF BANDUNG

    International Nuclear Information System (INIS)

    Evaluation of cooling instrumentation system of Triga Mark II reactor has been done. The reactor has been upgraded from 1 MW to 2 MW. The increasing of power is performed by changing the reactor components and systems. The reactor cooling system has important role in reactor operation, the system transfers heat produced in the core. The operation of the cooling system needed to be back up with qualified instrumentation. Evaluation has been done by doing analysis and observing the equipment design, type and clarification, performance study of instrumentation and system related to cooling system. It is known that the performance and system of Triga mark II reactor included the cooling system. It is also obtained the characteristic data of primary and secondary cooling system, piping diagram and instrumentation, emergency core cooling system. The cooling system has 4 measurement, i.e. flow rate, input and output temperature to heat exchanger, and electricity conductivity of water. The measurement can be observed from the reactor console. From this evaluation it is concluded that cooling system instrumentation followed the required criteria

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

    International Nuclear Information System (INIS)

    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

  12. Vaporization Rate Analysis of Primary Cooling Water from Reactor PUSPATI TRIGA (RTP) Tank

    International Nuclear Information System (INIS)

    Primary cooling system consists of pumps, heat exchangers, probes, a nitrogen-16 diffuser and associated valves is connected to the reactor TRIGA PUSPATI (RTP) tank by aluminium pipes. Both the primary cooling system and the reactor tank is filled with demineralized light water (H2O), which serves as a coolant, moderator as well as shielding. During reactor operation, vaporization in the reactor tank will reduce the primary water and contribute to the formation of vapor in the reactor hall. The vaporization may influence the function of the water subsequently may affect the safety of the reactor operation. It is essential to know the vaporization rate of the primary water to ensure its functionality. This paper will present the vaporization rate of the primary cooling water from the reactor tank and the influence of temperature of the water in the reactor tank to the vaporization rate. (author)

  13. Use of natural circulation mechanism core cooling of high temperature helium-cooled reactors as a means of safety enhancement

    International Nuclear Information System (INIS)

    Attempts were made to establish operating limits of the natural circulation mechanism as applied to nuclear power plants with gas-cooled thermal and fast reactors where helium is used as coolant. For this purpose parametric analysis is carried out for a closed loop with the reactor core as source of coolant heating and cooler (shutdown heat exchanger, steam generator or high-temperature heat exchanger) placed above the level of the upper end of the core. The gas moves due to its various density in the circuit riser and downcomer. The analysis made it possible to conclude that among numerous measures envisaged in design of gas-cooled nuclear reactors to provide safe operation, the natural circulation mechanism can be considered as one of the reliable and simple means of reactor core cooling except for depressurization case

  14. Fission product revaporization in the reactor cooling system

    International Nuclear Information System (INIS)

    The reactor cooling system (RCS) of an LWR can act as an efficient scrubber of volatile fission products released during a meltdown accident before vessel melt-through. This assertion is based on calculations that consider transport of the volatile fission products as vapours or condensed on particles. Retention in the primary system occurs by condensation or reaction with structural surfaces or by fallout of particles containing fission products. It is shown that this picture is perturbed by inclusion of decay heating in the thermal-hydraulic calculations. To do so we make use of the TRAP-MELT3 code which integrates the MERGE and TRAP-MELT2 codes and thus permits simultaneous calculation of thermal-hydraulics and fission product transport in the RCS during the meltdown phase of a severe LWR accident. Calculations on the Surry TMLB' sequence show that while structure temperatures can rise as much as 100 K with inclusion of decay heat, little additional fission product release from the RCS results before melt-through of the reactor vessel. After melt-through, structural temperatures are likely to continue to rise and fission products migrate along the RCS by revolatilizing in the hotter regions and condensing in the cooler regions. The potential for a significant source term of volatile fission products to the containment after melt-through thus exists. For these materials, therefore, the RCS may act more as a retardant than a retainer. Quantification of this conjecture will require further analyses. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-09-01

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

  17. Modelling containment passive safety systems in advanced water cooled reactors

    International Nuclear Information System (INIS)

    Most designs of advanced passive reactors incorporate Passive Containment Cooling Systems (PCCS) relying on steam condensation to cope with possible pressure increase that would result in the case of a postulated accident. As a consequence, experimental and analytical research programmes have been launched worldwide to investigate new configurations and conditions involved in these new scenarios. This paper summarises the major outcomes of the joint research of CIEMAT, UPV, and UW in developing predictive models to address anticipated conditions in the Simplified Boiling Water Reactors (CIEMAT-UPV) and in the AP600 (CIEMAT-UW). Even though both models share some of their fundamental characteristics (such as being mass/heat transfer analogy based), samples of their validation against independent databases illustrate their intrinsic differences in formulation according to the scenarios addressed by each one. Relative importance of condensate film or gas mixture velocity are discussed, and the effect of key factors such as noncondensable gas presence and pressure are stated. Experimental data from University of Berkeley (UCB) and from University of Wisconsin - Madison (UW) will be used to support comparisons and discussions held in the paper. In short, this work demonstrates that heat/mass transfer analogy-based models, particularly those relying on diffusion film modelling to account for noncondensable gas presence, are extremely useful in test interpretation and result in good agreement with reliable databases. (author)

  18. Uncertainties in analysis of innovative lead-cooled fast reactors

    International Nuclear Information System (INIS)

    There are numerous uncertainties in the prediction of innovative reactor design, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) core benchmark and the new cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and i designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include MOCUP-a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on ENDF/B-VI evaluations; and KWO2-a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 groups library. Uncertainties in the cross sections of lead were found particularly large and deserve careful evaluation. (author)

  19. Thermohydraulics of emergency core cooling in light water reactors

    International Nuclear Information System (INIS)

    This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided

  20. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steven J. Piet; Samuel E. Bays; Michael A. Pope; Gilles J. Youinou

    2010-11-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  1. Chemistry control strategies for a supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    The long-term viability of any Generation IV Supercritical Water-cooled Reactor (SCWR) concept depends on the ability of reactor designers and operators to predict and control water chemistry to minimize corrosion and corrosion product transport. Currently, SCWR material testing is being carried out using a limited range of water chemistries to establish the dependencies of the corrosion behavior on parameters such as water temperature and dissolved oxygen concentration. Once a final suite of candidate alloys is identified, more detailed, longer term testing will be needed under water chemistry regimes expected to be used in the SCWR. Prior to these tests, it will be necessary to identify proposed water chemistry regimes for the SCWR, and provide expected ranges for the key parameters. The direct-cycle design proposed for various SCWR concepts take advantage of the extensive operating experience world-wide of fossil-fired SCW power plants. Conceptually, the SCWR replaces the fossil-fired boiler with the reactor core (pressure vessel or pressure tube); the concept is broadly similar to that of a boiling water reactor. Current fossil-fired SCW power plants use either an all-volatile treatment or oxygenated water treatment for the feedwater systems. While the optimal water chemistry for a SCWR is yet to be determined, the monitored parameters are likely to be the same as those in existing fossil-fired and nuclear power plants: pH; conductivity, and concentrations of O2, H2, additives, impurities, corrosion and activation products. Monitoring will be required at many of the same components: feedwater, main 'steam', drains, pump outlets, condenser hotwell, and purification inlets and outlets. This paper outlines the strategy being used to develop a water chemistry regime for a CANDU® SCWR. It describes the key areas identified for chemistry monitoring and control: a) the reactor core, where materials are subjected to irradiation and high temperature, b

  2. Solar cooling system. Cooling with solar energy and hot air. Solare Kuehlanlage. Kuehlen mit Sonne und heisser Luft

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1999-06-01

    Engineers of Stuttgart Technical University are working on a solar cooling system with air-filled solar collectors. The first of these systems will start operation in July in a production hall in the Calw district. It will be Europe's first commercially used system for cooling and heating on the basis of solar energy and hot air.

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

    International Nuclear Information System (INIS)

    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

  4. High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

    International Nuclear Information System (INIS)

    Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies

  5. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

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

  6. A fundamentally new approach to air-cooled heat exchangers.

    Energy Technology Data Exchange (ETDEWEB)

    Koplow, Jeffrey P.

    2010-01-01

    We describe breakthrough results obtained in a feasibility study of a fundamentally new architecture for air-cooled heat exchangers. A longstanding but largely unrealized opportunity in energy efficiency concerns the performance of air-cooled heat exchangers used in air conditioners, heat pumps, and refrigeration equipment. In the case of residential air conditioners, for example, the typical performance of the air cooled heat exchangers used for condensers and evaporators is at best marginal from the standpoint the of achieving maximum the possible coefficient of performance (COP). If by some means it were possible to reduce the thermal resistance of these heat exchangers to a negligible level, a typical energy savings of order 30% could be immediately realized. It has long been known that a several-fold increase in heat exchanger size, in conjunction with the use of much higher volumetric flow rates, provides a straight-forward path to this goal but is not practical from the standpoint of real world applications. The tension in the market place between the need for energy efficiency and logistical considerations such as equipment size, cost and operating noise has resulted in a compromise that is far from ideal. This is the reason that a typical residential air conditioner exhibits significant sensitivity to reductions in fan speed and/or fouling of the heat exchanger surface. The prevailing wisdom is that little can be done to improve this situation; the 'fan-plus-finned-heat-sink' heat exchanger architecture used throughout the energy sector represents an extremely mature technology for which there is little opportunity for further optimization. But the fact remains that conventional fan-plus-finned-heat-sink technology simply doesn't work that well. Their primary physical limitation to performance (i.e. low thermal resistance) is the boundary layer of motionless air that adheres to and envelops all surfaces of the heat exchanger. Within this

  7. Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

    International Nuclear Information System (INIS)

    One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

  8. Advanced water-cooled reactor technologies. Rationale, state of progress and outlook

    International Nuclear Information System (INIS)

    Eighty per cent of the world's power reactors are water cooled and moderated. Many improvements in their design and operation have been implemented since the first such reactor started commercial operation in 1957. This report addresses the safety, environmental and economic rationales for further improvements, as well as their relevance to currently operating water reactors

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

    International Nuclear Information System (INIS)

    The MARS nuclear plant is a 600 MWth PWR with completely passive core safeguards. The most relevant innovative safety system is the Emergency Core Cooling System (ECCS), which is based on natural circulation, and on a passive-type activation that follows a core flow decrease, whatever was the cause (only one component, 400% redundant, is not static). The main thermal hydraulic transients occurring as a consequence of design basis accidents for the MARS plant were presented at the ICONE 3 Conference. Those transients were analyzed in the first stage, with the aim at pointing out the capability of the innovative ECCS to intervene. So, they included only a short-time analysis (extended for a few hundreds of seconds) and the well known RELAP 5 computer program was used for this purpose. In the present paper, the long-term analyses (extended for several thousands of seconds) of the same transients are shown. These analyses confirmed that the performance of the Emergency Core Cooling System of the MARS reactor is guaranteed also in long-term scenarios

  10. Air quality impact analysis in support of the new production reactor environmental impact statement

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory (PNL) conducted this air quality impact analysis for the US Department of Energy (DOE). The purpose of this work was to provide Argonne National Laboratory (ANL) with the required estimates of ground-level concentrations of five criteria air pollutants at the Hanford Site boundary from each of the stationary sources associated with the new production reactor (NPR) and its supporting facilities. The DOE proposes to provide new production capacity for the primary production of tritium and secondary production of plutonium to support the US nuclear weapons program. Three alternative reactor technologies are being considered by DOE: the light-water reactor, the low-temperature, heavy-water reactor, and the modular high-temperature, gas-cooled reactor. In this study, PNL provided estimates of the impacts of the proposed action on the ground-level concentration of the criteria air pollutants for each of the alternative technologies. The criteria pollutants were sulfur dioxide, nitrogen dioxide, carbon monoxide, total suspended particulates, and particulates with a diameter of less than 10 microns. Ground-level concentrations were estimated for the peak construction phase activities expected to occur in 1997 and for the operational phase activities beginning in the year 2000. Ground-level concentrations of the primary air pollutants were estimated to be well below any of the applicable national or state ambient air quality standards. 12 refs., 19 tabs

  11. Analysis of Steam Condensation in a Finned Tube of Air-Water Combined Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Taesoon; Bae, S. W.; Kim, K. H.; Park, Y. S.; Park, H. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    A Passive Auxiliary Feedwater System (PAFS) is one of the passive cooling systems of the existing power plant and the operating period is 8 hours because of the limited capacity of the cooling water tank. Therefore, to increase the operating period from 8 to 72 hours for an existing PAFS, the capacity of the cooling water tank should be increased up to 3-4 times. To resolve the excessive increase of the cooling tank volume in water cooling systems, an air-water combined passive cooling system is proposed. In this combined cooling system, the core cooling during the initial stage of an accident having high decay power depends on the water cooling systems such as PAFS. For the later phase of an accident, an air-cooling system is applied to the core cooling. In the operation of the air-cooling system, the steam from the cooling water tank of the PAFS is condensed and recirculated to the cooling water tank by an air-cooling heat exchanger. In this way, the increase of the cooling water tank volume can be minimized. To design an air-water combined cooling system, the steam condensation characteristics of an air-cooling heat exchanger tube was calculated using a CFX code. The results show that the air velocities around the tube at the steam inlet/outlet regions are quite different with each other. Therefore, dense installation of thermocouples at the tube bottom region is required to measure the steam condensation in the tube. Otherwise, the detection and measurement of steam condensation at the steam inlet region may be very difficult. The velocity distribution of air is not uniform and the distributions of air temperature and velocity around the heat exchanger tube are strongly asymmetric. In the design of the measurement system of the test facility, the problems mentioned above should be considered.

  12. Analysis of Steam Condensation in a Finned Tube of Air-Water Combined Cooling System

    International Nuclear Information System (INIS)

    A Passive Auxiliary Feedwater System (PAFS) is one of the passive cooling systems of the existing power plant and the operating period is 8 hours because of the limited capacity of the cooling water tank. Therefore, to increase the operating period from 8 to 72 hours for an existing PAFS, the capacity of the cooling water tank should be increased up to 3-4 times. To resolve the excessive increase of the cooling tank volume in water cooling systems, an air-water combined passive cooling system is proposed. In this combined cooling system, the core cooling during the initial stage of an accident having high decay power depends on the water cooling systems such as PAFS. For the later phase of an accident, an air-cooling system is applied to the core cooling. In the operation of the air-cooling system, the steam from the cooling water tank of the PAFS is condensed and recirculated to the cooling water tank by an air-cooling heat exchanger. In this way, the increase of the cooling water tank volume can be minimized. To design an air-water combined cooling system, the steam condensation characteristics of an air-cooling heat exchanger tube was calculated using a CFX code. The results show that the air velocities around the tube at the steam inlet/outlet regions are quite different with each other. Therefore, dense installation of thermocouples at the tube bottom region is required to measure the steam condensation in the tube. Otherwise, the detection and measurement of steam condensation at the steam inlet region may be very difficult. The velocity distribution of air is not uniform and the distributions of air temperature and velocity around the heat exchanger tube are strongly asymmetric. In the design of the measurement system of the test facility, the problems mentioned above should be considered

  13. Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

    International Nuclear Information System (INIS)

    Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 233U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

  14. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

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

  17. High temperature gas-cooled reactors - Operating on fuel recycle

    International Nuclear Information System (INIS)

    The HTGR, because of a unique combination of design characteristics, is a resource-efficient and cost-effective reactor. In the HTGR, the low power-density core, coated particle fuel design, and gas cooling combine to provide high neutron economy, fuel burnup and thermodynamic efficiency. Under recycle uranium assumptions, the resource utilization is particularly attractive due to the high neutronic value of the bred, and recycled, U-233 produced from the thorium irradiation. The uranium resource requirements for the current MEU/Th cycle with annual refueling results in a 30-year net U3O8 requirement of 3030 ST/GWe. The basic design of the HTGR refueling scheme, whereby only selected regions of the core need be accessible during each refueling, makes fuel utilization improvements through semi-annual refueling an acceptable alternative in terms of plant availability. This alternative reduces the 30-year U3O8 requirement by about 10%. Additional resource utilization improvements of 11 to 14% could be realized by improved fuel management techniques

  18. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    International Nuclear Information System (INIS)

    Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept. The objective of the proposed research was to investigate degradation of materials in the supercritical water environment (SCW). First, representative alloys from the important classes of candidate materials were studied for their corrosion and stress-corrosion cracking (SCC) resistance in supercritical water. These included ferritic/martensitic (F/M) steels, austenitic stainless steels, and Ni-base alloys. Corrosion and SCC tests were conducted at various temperatures and exposure times, as well as in various water chemistries. Second, emerging plasma surface modification and grain boundary engineering technologies were applied to modify the near surface chemistry, microstructure, and stress-state of the alloys prior to corrosion testing. Third, the effect of irradiation on corrosion and SCC of alloys in the as-received and modified/engineered conditions were examined by irradiating samples using high-energy protons and then exposing them to SCW

  19. Development Project of Supercritical-water Cooled Power Reactor

    International Nuclear Information System (INIS)

    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)

  20. Hydrogen in water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    The Commission of the European Community (CEC) and the International Atomic Energy Agency (IAEA) decided in 1989 to update the state of the art concerning hydrogen in water cooled nuclear power reactors by commissioning a report which would review, all the available information to-date and make recommendations for the future. This joint report was prepared by committees formed by the IAEA and by the CEC. The aim of this report is to review the current understanding on the areas in which the research on hydrogen in LWR is conventionally presented, taking into account the results of the latest reported research developments. The main reactions through which hydrogen is produced are assessed together with their timings. An estimation of the amount of hydrogen produced by each reaction is given, in order to reckon their relative contribution to the hazard. An overview is then given of the state of knowledge of the most important phenomena taking place during its transport from the place of production and the phenomena which control the hydrogen combustion and the consequences of combustion under various conditions. Specific research work is recommended in each sector of the presented phenomena. The last topics reviewed in this report are the hydrogen detection and the prevent/mitigation of pressure and temperature loads on containment structures and structures and safety related equipment caused by hydrogen combustion

  1. Optimal design study of cylindrical finned reactor for solar adsorption cooling machine

    Energy Technology Data Exchange (ETDEWEB)

    Allouache, N. [Univ. des Sciences et de la Technologie Houari Boumediene, Bab Ezzouar (Algeria). Faculte de Genie Mecanique et de Genie des Procedes; Al Mers, A. [Moulay Ismail Univ., Meknes (Morocco). Ecole National Superieure d' Art et Metiers

    2010-07-01

    Solid adsorption cooling machines use medium temperature industrial waste heat together with a renewable energy source, such as solar energy. The adsorption cooling machine consists of an evaporator, a condenser and a reactor containing a solid adsorbent. In this study, a model was developed for thermodynamic performance analysis and optimization of a cylindrical finned solar reactor in an adsorption refrigerator working with activated carbon-ammonia. The heat and mass transfer in the adsorption cooling machine was determined. The model was validated using experimental results. The study investigated the sensitivity of the machine performance versus the geometrical configuration of the reactor. The study showed that for an optimized reactor, a higher fin number significantly reduces the heat losses of the reactor. It was concluded that the solar coefficient of performance (COP) of an optimized reactor can reach 45 per cent when the number of fins varies between 5 and 6. 10 refs., 4 figs.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1999. It was included into the IAEA's Programme following endorsement in 1995 by the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote international information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. (authors)

  4. Design measures in evolutionary water cooled reactors to optimize for economic viability

    International Nuclear Information System (INIS)

    Since the mid 1980s, there have been various efforts to develop evolutionary water cooled reactors based on the current operating plant experience. To sustain and improve the economic viability, particular attention has been paid to the following aspects in developing evolutionary water cooled reactors: design simplification and increased operating margins, standardization in design as well as construction and operation, integration of operating plant insights, and consideration of safety, operability and constructability during the design stage. This paper reviews each item and discusses several examples from some of the evolutionary water cooled reactors being developed. (author)

  5. Operational experiences with solar air collector driven desiccant cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Eicker, Ursula; Schneider, Dietrich; Schumacher, Juergen [Faculty of Civil Engineering, Building Physics and Economics, University of Applied Sciences Stuttgart, Schellingstrasse 24, D-70174 Stuttgart (Germany); Ge, Tianshu; Dai, Yanjun [Faculty of Civil Engineering, Building Physics and Economics, University of Applied Sciences Stuttgart, Schellingstrasse 24, D-70174 Stuttgart (Germany); Institute of Refrigeration and Cryogenics, Solar Energy Research Center, Shanghai Jiao Tong University, Shanghai 200240 (China)

    2010-12-15

    Component performance and seasonal operational experiences have been analysed for desiccant cooling systems powered by solar air collectors. Measurements during the commissioning phase in Spain (public library) and in Germany (production hall) showed that the dehumidification efficiency of the sorption rotors was 80% and the humidification efficiency of the contact evaporators was 85-86%. Only in a two-stage desiccant system monitored in China (laboratory building), a dehumidification efficiency of 88% was reached. The rotary heat exchangers only had 62-68% measured heat recovery efficiency, which is lower than specified. Seasonal performance monitoring carried out in the German installation showed that average seasonal COP's were close to 1.0, when related to all operation hours. COP's increase if low regeneration temperatures are used with low dehumidification rates, which is often sufficient for moderate German climatic conditions, but much less so in the humid Chinese climate. Electrical COP's for the German system including air distribution were between 1.7 and 4.6 and reach values of 7.4, when only additional pressure drops of the desiccant unit are considered. It could be shown that conventional control strategies lead to high auxiliary energy consumption, for example if fixed heating setpoint temperatures are used. Furthermore the solar air collector energy yield was very low in the German system, as regeneration was only used when all other options such as humidification at high air volume flows did not reduce the room air temperature enough. The studies showed that the measured auxiliary energy consumption could be reduced to near zero, if regeneration temperature setpoints were not fixed to constant values. The solar air collector efficiency was good at about 50% both for the flat plate collectors used in Spain and Germany and the Chinese vacuum tube solution. A cost analysis demonstrated the viability of the concept, if some funding of

  6. Loss of coolant accident analysis of supercritical water-cooled reactor fuel qualification test loop

    International Nuclear Information System (INIS)

    The supercritical water-cooled reactor fuel qualification test (SCWR-FQT) intends to test a small scale fuel assembly under supercritical water environment in a research reactor. The modified ATHLET code was applied to model the supercritical water-cooled experimental loop containing this fuel assembly and to perform the calculation analysis of the loss of coolant accident induced by the coolant pipe break. The results indicate that the design of existing safety system can practically ensure the effective cooling of the fuel rod experimental section in the accident scenario. The results also show that the modified ATHLET code has good suitability in simulation of supercritical water-cooled system. (authors)

  7. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    International Nuclear Information System (INIS)

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li4SiO4, Li2TiO3 and LiAlO2 compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year

  8. Performance of high-temperature gas-cooled reactor as a tritium production device for fusion reactors

    International Nuclear Information System (INIS)

    Highlights: ► The performance of a gas-cooled reactor as a tritium production device was studied. ► Gas-cooled reactors with 3 GWt output can produce 5–8 kg of tritium in a year. ► Use of Li2O compound is efficient compared with Li4SiO4 or Li2TiO3 one. ► Amount of tritium produced can be increased by reducing the enrichment of 235U. - Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production device is examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target of a typical gas-cooled reactor, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the 3-dimensional entire-core region of GTHTR300 were carried out considering its unique double heterogeneity structure. It is shown that gas-cooled reactors with thermal output power of 3 GW in all can produce 5–8 kg of tritium in a year.

  9. CFD analysis of decay heat removal scenarios of the lead cooled ELSY reactor

    International Nuclear Information System (INIS)

    The lead cooled European reactor concept ELSY is characterized by its innovative, compact design, where all components of the primary loops are located inside the reactor vessel. The vessel includes 8 steam generators and pumps which generate a coolant flux of 126 tons/s. At nominal operation conditions the core releases about 1500 MW converted to an electric power of 600 MW. For the coolant temperatures 400 C at the core inlet and 480 C are envisaged. If the reactor is shut down and the pumps are switched off, the decay heat is removed by isolation condenser (IC) systems acting on the SG secondary circuits. If the IC's are not available, heat removal by 4 dip coolers is foreseen, which can be operated by gravitation driven water flow or by air. Additionally the outer vessel wall is permanently cooled by a RVACS (reactor vessel air cooling) system located between the outer vessel wall and the reactor cavity. The main intention of this work is the investigation of the passive cooling systems which are used for the decay heat removal. The CFD vessel model of about 20 million cells takes advantage of the components symmetry and simulates a 90 deg. section of the reactor. The spatial resolution of the computational grid varies between 5 mm close to walls and 100 mm in undisturbed regions where only small gradients are expected. The core, the pumps and the SG's are simulated with porous media models including volumetric source terms for momentum and energy. For the SG's detailed CFD studies with at least one order of magnitude finer grid resolution are performed in order to obtain data for pressure losses. Solids like pipe walls or the core barrel are taking into account by heat conduction. As the model considers a closed system the coolant flux is controlled by momentum sources of the pumps and frictional losses mainly by the core and the SG's. The simulations are performed as single phase flows, therefore the free lead surface at the unclosed upper part of the vessel

  10. Technical assessment of gas turbine cycle for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The gas turbine cycle appears to be the best near-term power conversion method for the high temperature gas-cooled reactor (HTGR). The author extensively investigates the gas turbine cycle including direct cycle, open indirect cycle and closed indirect cycle with medium of helium, nitrogen and air. Each cycle is analyzed and optimized from the thermodynamic standpoint and its turbo-machine is aerodynamically designed. As a result, the direct cycle with helium is an ideal option for the HTGR gas turbine cycle; however it is not easy to be realized based on current technology. The closed indirect cycle with helium or nitrogen is a practical one at present time, which can get the gas turbine cycle and lay technical bases for the future direct cycle

  11. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  12. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    International Nuclear Information System (INIS)

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution

  13. Water chemistry management in cooling system of research reactor in JAERI

    International Nuclear Information System (INIS)

    The department of research reactor presently operates three research reactors (JRR-2, JRR-3M and JRR-4). For controlling and management of water and gas in each research reactor are performed by the staffs of the research reactor technology development division. Water chemistry management of each research reactor is one of the important subject. The main objects are to prevent the corrosion of water cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the radioactive waste. In this report describe a outline of each research reactor facilities, radiochemical analytical methods and chemical analytical methods for water chemistry management. (author)

  14. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  15. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    International Nuclear Information System (INIS)

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  16. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  17. Numerical study by large-eddy simulation on effects and mechanism of air-cooling enhancing technologies

    International Nuclear Information System (INIS)

    Learning from the lessons of the Fukushima Daiichi nuclear incident in which a long-term station black-out occurred, we have been developing an air-cooling system for boiling water reactors that can operate without electricity for a virtually indefinite time. Improvement in the heat transfer performance of air-cooling is key to the development of the air-cooling system. We developed air-cooling enhancing technologies for the air-cooling system by using heat transfer fins, turbulence-enhancing ribs and a micro-fabrication surface. In our previous study, the performance of these air-cooling enhancing technologies was evaluated by heat transfer tests using a single pipe of the air-cooling heat exchanger. To achieve further improvement of the heat transfer performance, it is important to understand the mechanism of the air-cooling enhancing technologies. In this study, we used the numerical analysis which is based on the filtered incompressible Navier-Stokes equation and the filtered energy equation with the large-eddy simulation in order to investigate the effects and the mechanism of the developed air-cooling enhancing technologies. We found that the analysis results agreed well with the experimental results and the empirical formula results. The heat transfer enhancement mechanism of the heat transfer fin is due to an increase in the heat transfer area. Due to a decrease in the flow velocity at the base of the fins, the increase in the Nusselt number was approximately 15% smaller than the estimated value from the area increase. In the heat transfer enhancement by the turbulence-enhancing ribs, the unsteady behavior of the large-scale vortex generated by the flow separation plays an important role. The enhancement ratio of the Nusselt number by the micro-fabrication surface can be explained by the apparent thermal conductivity. The Nusselt number was increased 4-8% by the micro-fabrication surface. The effect of the micro-fabrication surface is increased by applying

  18. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    International Nuclear Information System (INIS)

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic (∼47%), wood (∼38%) and asbestos transite (∼14%). The remaining ∼1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste, except for the

  19. Status of TRIOU code for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Highlights: ► The paper reports the recent improvements of TRIOU CFD code for SFR applications. ► Subassembly refined modelling and sub-channel modelling applications are presented. ► A global core modelling including the inter-wrapper region and the connection to the hot plenum is described. ► Special development of Front Tracking method is shown to deal with free surface and gas entrainment. ► For decay heat removal situations, the whole primary vessel is modelled with TRIOU code. - Abstract: Sodium cooled fast reactors (SFRs) have been developed in France for nearly 50 years with successively Rapsodie, Phenix and Superphenix plants. Nowadays, the so-called Astrid prototype is developed in France in the frame of Generation IV deployment. The Astrid project requires thermal hydraulic inputs to support the design and the safety analysis. This paper deals with some thermal hydraulic concerns in the primary circuit: the subassembly, the core, the hot plenum and the cold plenum. The so-called TRIOU Computational Fluid Dynamic (CFD) code developed at CEA has been progressively adapted to these Astrid concerns. The paper presents the recent improvements, the present status and the remaining challenges for TRIOU code on each topic. For the subassembly, refined modelling and sub-channel modelling have been developed in parallel. The validation process based on existing experimental data is in progress. A global core modelling including the inter-wrapper region and the connection to the hot plenum is depicted. The need of experimental validation is pointed out. The core outlet region requires refined Large Eddy Simulation computations to predict temperature fluctuations which can induce thermal fatigue. Validation based on sodium experimental data is briefly presented. Thermal stratification in the plenum is a key point for thermal stress analysis on the structures. Validation process includes the comparison to reactor data. Special developments using a Front

  20. Neutronics of a liquid salt cooled - very high temperature reactor

    International Nuclear Information System (INIS)

    During last few years, the interest in the innovative, Liquid Salt cooled - Very High Temperature Reactor (LS-VHTR), has been growing. The preconceptual design of the LS-VHTR was suggested in Oak Ridge National Laboratory (ORNL) [1] and nowadays, several research institutions contribute to the development of this concept. The LS-VHTR design utilises a prismatic, High Temperature Reactor (HTR) fuel [2] in combination with liquid salt as a coolant. This connection of high-performance fuel and a coolant with enhanced heat transfer abilities enables efficient and economical operation. Main objective of the LS-VHTR operation may be either an efficient electricity production or a heat supply for a production of hydrogen or, combination of both. The LS-VHTR is moderated by graphite. The graphite matrix of the fuel blocks, as well as the inner and outer core reflectors serve as a thermal buffer in case of an accident, and they provide a strong thermal feedback during normal reactor operation. The high inherent safety of the LS-VHTR meets the strict requirements on future reactor systems, as defined by the Gen IV project. This work, purpose, scope, contribution to the state-of-art: The design, used in the present work is based on the first ORNL suggestion [1]. Recent study is focused on comparison of the neutronic performance of two types of fuel in the LS-VHTR core, whereas, in all previous works, only uranium fuel has been investigated. The first type of fuel, which has been employed in the present analysis, is based on the spent Light Water Reactor (LWR) fuel, whereas the second one consists of enriched uranium oxide. The results of such a comparison bring a valuable knowledge about limits and possibilities of the LS-VHTR concept, when employed as a spent fuel burner. Method:It is used a 3-D drawing of the LS-VHTR core, which contains 324x10 hexagonal fuel blocks. Each fuel block contains 216x10 fuel pins, which consists of TRISO particles incorporated into a graphite

  1. Conceptual Design of Primary Cooling System for an Advanced HANARO Reactor

    International Nuclear Information System (INIS)

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates a 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 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. Through this study, the conceptual design of the primary cooling system has been established including design requirements, performance requirements, design restrictions, system descriptions and system operation to maintain the system function. And preliminary design requirement of the primary cooling system has been established in based on the conceptual design

  2. An experimental investigation of the air entrainment in the shutdown cooling system during mid-loop operation

    International Nuclear Information System (INIS)

    An experimental study on the air entrainment phenomena during mid-loop operation has been performed for Ulchin 3 and 4 nuclear power plant (UCN 3 and 4). The UCN 3 and 4 is the standard Combustion Engineering (CE) System 80, two-loop, 2825 MWt pressurized water reactor which is currently under construction in Korea. This study was undertaken by Korea Atomic Energy Research Institute to provide a basis for modification of CE system 80 design which has relatively small mid-loop operating range, and to investigate the impact of the air entrainment on the shutdown cooling pump. A 1/4 scale model test was performed for two shutdown cooling suction nozzle configurations, with and without bellmouth entry, to obtain data relative to air ingestion and vortex formation at the shutdown cooling suction nozzle during mid-loop operation. The test model size was determined based on Froude number corresponding to that which occurs in plants by considering that the Weber number and Reynolds number are large enough to ensure that liquid surface tension and viscosity would not significantly affect the vortex formation. An empirical correlation between the flow rate and the critical submergence was obtained with respect to the Froude number. The impact of the air entrainment of the pump was also investigated to select parameters for plant operators to monitor the onset of air entrainment into shutdown cooling system (SCS). Effects of the bellmouth entry on the critical submergence were also investigated. (author)

  3. A proposal of cooling source on full power loss in nuclear reactors

    International Nuclear Information System (INIS)

    A cooling method is proposed of residual power released after the shutdown of nucleate reactors even when all electricity becomes unavailable. The cooling source is a water reservoir that the initial level is 20 m high from the sea. Only turbine driven pumps are used for circulation of coolant in the reactor. The high-pressure injection system is used to remove residual power directly from the reactor vessel, then cooled with the heat exchanger in the residual heat removal system, to which a turbine-driven pump in the reactor core isolation cooling system supplies water from the suppression pool. The flow system of cooling water consists of feeding pipe, tubes in the heat exchanger and an orifice that controls flow rate. The simulation reveals that the temperature of the suppression pool has its maximum at about 50 hours from the shutdown, though the flow rate is steadily decreasing. It is results from the decrease of residual power. The temperature increases again in the last phase of cooling because the flow rate of cooling water diminishes. The delay of the start of cooling contributes to make the cooling water exit temperature cooler at the last phase of cooling. Two restrictions were applied to minimize the area of the reservoir. One is that the temperature of the suppression pool does not exceed designed temperature of the containment vessel. The other is that the cooling water exit temperature is less than 60°C. The minimized area is 3094 m2, a possible area to build. This cooling system is considered very important for reactors located on seashore, where tsunami attack may destroy the residual heat removal sea water system. (author)

  4. High temperature corrosion of structural materials under gas-cooled reactor helium

    International Nuclear Information System (INIS)

    The Generation IV International Forum has selected six promising nuclear power systems for further collaborative investigations and development. Among these six concepts, two candidates are Gas Cooled Reactors (GCR), namely the Very High Temperature Reactor (VHTR) and the Gas-cooled Fast Reactor (GFR). The CEA has launched a R and D program on the metallic materials for application in an innovative GCR. Structural GCR alloys have been extensively studied in the past three decades. Some critical aspects for the steels and nickel base alloys resistance under the service conditions are microstructural stability, creep strength and compatibility with the cooling gas. The coolant, namely helium, proved to contain impurities mainly H2, CO, CH4, N2 and steam in the microbar range that interact with metals at high temperature. Surface scale formation, bulk carburisation and/or decarburisation can occur, depending on the atmosphere characteristics, primarily the effective oxygen partial pressure and carbon activity, on the temperature and on the alloys chemical composition. These structural transformations can notably influence the mechanical properties: carburisation may induce a loss in toughness and ductility whereas decarburisation impedes the creep strength. There is a valuable theoretical as well as practical knowledge on the corrosion of high temperature alloys in the primary circuit of a GCR but this past experience is not sufficient to qualify every component in a future reactor. On the one hand, the material environment could be significantly different from the former GCR's, especially regarding the higher temperature. On the other hand, the materials of interest are partly different. Ni-Cr-W alloys, for instance, may offer significant improvement in the maximum operating temperature as far as the mechanical properties are concerned. However, their corrosion resistance toward the GCR atmosphere is still unknown. We describe here our first corrosion tests of Haynes

  5. Air cooling of disk of a solid integrally cast turbine rotor for an automotive gas turbine

    Science.gov (United States)

    Gladden, H. J.

    1977-01-01

    A thermal analysis is made of surface cooling of a solid, integrally cast turbine rotor disk for an automotive gas turbine engine. Air purge and impingement cooling schemes are considered and compared with an uncooled reference case. Substantial reductions in blade temperature are predicted with each of the cooling schemes studied. It is shown that air cooling can result in a substantial gain in the stress-rupture life of the blade. Alternatively, increases in the turbine inlet temperature are possible.

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

    International Nuclear Information System (INIS)

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

  7. Gas-cooled fast breeder reactor shielding benchmark calculation

    Energy Technology Data Exchange (ETDEWEB)

    Rouse, C.A.; Mathews, D.R.; Koch, P.K.

    1977-01-01

    This report summarizes the results of a shielding benchmark calculation performed by General Atomic (GA) and Oak Ridge National Laboratory (ORNL). The problem analyzed was a neutron-coupled gamma ray transport calculation of the core blanket shield of the 300-MW(e) gas-cooled fast breeder reactor (GCFR). Comparison of the initial GA and ORNL results indicated good agreement for fast fluxes (E greater than 0.9 MeV and E greater than 0.086 MeV) but poor agreement for epithermal and thermal neutron fluxes. Examination of the results revealed that a deficiency in the GA fine-group cross section preparation code was responsible for the differences in the GA and ORNL iron cross sections. Modification of the GA cross sections to include self-shielding was accomplished, and the updated GA benchmark calculation performed with the self-shielded iron cross sections was in excellent agreement with the ORNL results for fast neutron fluxes with E greater than 0.9 MeV and E greater than 0.086 MeV and in good agreement for epithermal and thermal fluxes. The agreement of the gamma heating rates also improved significantly. Thus, it was concluded that the good agreement of the GA and ORNL neutron-coupled gamma ray transport calculation indicates that (1) the methods and cross sections used by both laboratories were compatible and consistent and (2) the use of 24 neutron energy groups and 15 gamma energy groups by GA was adequate compared with the use of 51 neutron energy groups and 25 gamma energy groups by ORNL.

  8. Gas entrainment issues in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Highlights: • Main sources of gas and related issues in SFR are presented. • Various approaches of gas transport are briefly described. • Previous experimental studies to reduce gas entrainment are reported. • Present evaluation of free surface gas entrainment is presented. - Abstract: Sodium cooled fast reactors have been developed in France for nearly 50 years. The so-called Astrid technology demonstrator is currently designed in the frame of Generation IV deployment. Gas entrainment in the primary sodium circuit is a key issue as it can lead to safety problems in case of accumulation and transport of large quantity of gas through the core. The paper first introduces the main problems caused by the presence of gas in the primary sodium circuit, the various sources of gas and the main issues on gas transport. As sodium–argon free surface is potentially an important source of gas entrainment in the primary circuit, we present the main results obtained in past experimental studies on vortex type gas entrainment at free surface. Water tests were performed in a simple flow condition to study the physical process of vortex occurrence and gas entrainment. Other water tests were performed in representative hot pool models at different scales to analyze similarity criteria. Moreover, design improvements and local devices were tested to avoid gas entrainment at the free surface. Nowadays, numerical tools are progressively used to estimate the risk of gas entrainment at the free surface. We present the methodology in progress to define local criteria on vortex occurrence and gas entrainment, and to apply these criteria to global calculations of the whole pool. A Front-Tracking method coupled to a Large Eddy Simulation approach is implemented in TRIOU code to compute free surface instabilities and vortex occurrence. Experimental data from the literature are used to validate the numerical approach and a new test facility called BANGA is in progress at CEA to

  9. Safety aspects of forced flow cooldown transients in Modular High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.

    1992-09-01

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

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

    International Nuclear Information System (INIS)

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

  13. Prediction of Air Flow and Temperature Distribution Inside a Yogurt Cooling Room Using Computational Fluid Dynamics

    OpenAIRE

    Surendhar, A.; V.M. Sivakumar; Kannadasan, T.

    2015-01-01

    Air flow and heat transfer inside a yogurt cooling room were analysed using Computational Fluid Dynamics. Air flow and heat transfer models were based on 3D, unsteady state, incompressible, Reynolds-averaged Navier-Stokes equations and energy equations. Yogurt cooling room was modelled with the measured geometry using 3D design tool AutoCAD. Yogurt cooling room model was exported into the flow simulation software by specifying properties of inlet air, yogurt, pallet and walls of the room. Pac...

  14. Critical experiments and reactor physics calculations for low enriched high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    On the recommendation of the IAEA International Working Group on Gas Cooled Reactors, the IAEA established a Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low-Enriched High Temperature Gas Cooled Reactors (HTGRs) in 1990. The objective of the CRP was to provide safety-related physics data for low-enriched uranium (LEU) fueled HTGRs for use in validating reactor physics codes used by the participating countries for analyses of their designs. Experience on low-enriched uranium, graphite-moderated reactor systems from research institutes and critical facilities in participating countries were brought into the CRP and shared among participating institutes. The status of experimental data and code validation for HTGRs and the remaining needs at the initiation of this CRP were addressed in detail at the IAEA Specialists Meeting on Uncertainties in Physics Calculations for HTGR Cores held at the Paul Scherrer Institute (PSI), Villigen, Switzerland in May, 1990. The main activities of the CRP were conducted within an international project at the PROTEUS critical experiment facility at the Paul Scherrer Institute, Villigen, Switzerland. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Fuel for the experiments was provided by the KFA Research Center, Juelich, Germany. Initial criticality was achieved on July 7, 1992. These experiments were conducted over a range of experimental parameters such as carbon-to-uranium ratio, core height-to-diameter ratio, and simulated moisture concentration. To assure that the experiments being conducted are appropriate for the design of the participants, specialists from each of the countries have participated

  15. 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)

    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

  16. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  17. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    International Nuclear Information System (INIS)

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days

  18. Mitigation of hydrogen hazards in water cooled power reactors

    International Nuclear Information System (INIS)

    Past considerations of hydrogen generated in containment buildings have tended to focus attention on design basis accidents (DBAs) where the extent of the in-core metal-water reaction is limited at low values by the operation of the emergency core cooling systems (ECCS). The radiolysis of water in the core and in the containment sump, together with the possible corrosion of metals and paints in the containment, are all relatively slow processes. Therefore, in DBAs the time scale involved for the generation of hydrogen allows sufficient time for initiation of measures to control the amount of hydrogen in the containment atmosphere and to prevent any burning. Provisions have been made in most plants to keep the local hydrogen concentration below its flammability limit (4% of volume) by means of mixing devices and thermal recombiners. Severe accidents, involving large scale core degradation and possibly even core concrete interactions, raise the possibility of hydrogen release rates greatly exceeding the capacity of conventional DBA hydrogen control measures. The accident at Three Mile Island illustrated the potential of unmitigated hydrogen accumulation to escalate the potential consequences of a severe accident. In a severe accident scenario, local high hydrogen concentrations can be reached in a short time, leading to flammable gas mixtures in containment. Another possibility is that local high steam concentrations will initially create an inert atmosphere and prevent burning for a limited time. While such temporary inerting provides additional time for mixing (dilution) of the hydrogen with containment air, depending on the quantity of hydrogen released, it prevents early intervention by deliberate ignition and sets up conditions for more severe combustion hazards after steam condensation eventually occurs, e.g., by spray initiation or the long term cooling down of the containment atmosphere. As the foregoing example indicates, analysis of the hydrogen threat in

  19. Prediction of nongaseous fission products behavior in the primary cooling system of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    In high temperature gas-cooled reactors (HTGRs), some amounts of fission products (FPs) are released mainly from fuel with failed coatings and are transported in the primary cooling system with the primary coolant during normal operation. In that case, condensable FPs plateout on the inner surface of components in the primary cooling system. On the other hand, since the HTGRs use helium gas as primary coolant, the primary coolant is not activated itself and very small amount of corrosion products is generated. Then, γ-ray emitted from the FPs becomes main source in shielding design of the HTGRs, and not only release amount from fuel but also plateout distributions of the FPs should be properly evaluated. Therefore, prediction of plateout behavior in the primary cooling system of HTGRs was carried out based on the calculation result of plateout distribution in High Temperature Engineering Test Reactor. Before the calculation, analytical model was verified by comparison with experimentally obtained plateout distributions and the applicability of the model to predict the plateout distributions in the primary cooling system of HTGR was certified. This report describes the predicted result of plateout distribution in the primary cooling system of HTGR together with the verification result of the analytical model. (author)

  20. Experimental Analysis Of 80 Tr Capacity Air Cooled Scroll Chiller Using R-22 & R-407c.

    Directory of Open Access Journals (Sweden)

    Mr. Bhikhu B,

    2014-04-01

    Full Text Available In air conditioning systems, chilled water is typically distributed to heat exchangers, or coils, in air handling units or other types of terminal devices which cool the air in their respective space(s, and then the water is recirculated back to the chiller to be cooled again. These cooling coils transfer sensible heat and latent heat from the air to the chilled water, thus cooling and usually dehumidifying the air stream. The experiment works on 80 TR capacity Air Cooled Scroll Chiller systems. Vapour compression refrigeration cycle is used for cooling chilling water. Capacity of compressor is taken same for the different refrigerants used for experimental analysis. During experimental work used R-22 and R-407C as refrigerants. Theoretical COP of system with R- 22 refrigerant is 4.166 and actual COP is 2.227. For 80 TR capacity scroll air cooled chiller, theoretical COP of system with R-407c refrigerant is 3.465 and actual COP is 2.745 respectively. Based on the result analysis the same capacity of air cooling system with scroll compressor Actual COP of R-407C is higher than R-22. It means R-407C is also a alternative refrigerant for air cooled chilling system and also for HVAC system.

  1. Simulation and Optimization of Air-Cooled PEMFC Stack for Lightweight Hybrid Vehicle Application

    OpenAIRE

    Jingming Liang; Zefeng Wu

    2015-01-01

    A model of 2 kW air-cooled proton exchange membrane fuel cell (PEMFC) stack has been built based upon the application of lightweight hybrid vehicle after analyzing the characteristics of heat transfer of the air-cooled stack. Different dissipating models of the air-cooled stack have been simulated and an optimal simulation model for air-cooled stack called convection heat transfer (CHT) model has been figured out by applying the computational fluid dynamics (CFD) software, based on which, the...

  2. Measures against the adverse impact of natural wind on air-cooled condensers in power plant

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The natural wind plays disadvantageous roles in the operation of air-cooled steam condensers in power plant.It is of use to take various measures against the adverse effect of wind for the performance improvement of air-cooled condensers.Based on representative 2×600 MW direct air-cooled power plant,three ways that can arrange and optimize the flow field of cooling air thus enhance the heat transfer of air-cooled condensers were proposed.The physical and mathematical models of air-cooled condensers with various flow leading measures were presented and the flow and temperature fields of cooling air were obtained by CFD simulation.The back pressures of turbine were calculated for different measures on the basis of the heat transfer model of air-cooled condensers.The results show that the performance of air-cooled condensers is improved thus the back pressure of turbine is lowered to some extent by taking measures against the adverse impact of natural wind.

  3. Education and Training in Support of Sodium Cooled Fast Reactors Around the World

    International Nuclear Information System (INIS)

    The Generation IV Technology Roadmap has identified six systems for their potential to meet the new technology goals to improve safety, sustainability, economic competitiveness and proliferation resistance. Among these systems, three are fast neutron reactors: two cooled by liquid metal, the sodium cooled fast reactor (SFR) and the lead cooled fast reactor, and one cooled by gas, the gas cooled fast reactor. The SFR has the most comprehensive technological basis as a result of the experience gained from worldwide operation of several experimental, prototype and commercial size reactors from the 1940s. In order to support the operation of existing reactors, design activities for new projects and decommissioning of old reactors, it is mandatory to maintain and develop skills, particularly among the young generation. This paper presents the current strategies developed at the national level, or within a multilateral framework such as the EU or the IAEA, to support the development of SFRs, with particular focus on education and training initiatives dedicated to students, researchers, designers and operators involved in the development of SFRs. (author)

  4. U.S. ALMR sodium cooled reactor design and performance

    International Nuclear Information System (INIS)

    The U.S. Advanced Liquid Metal Reactor (ALMR) has innovative features for reliable operation, accident prevention and severe accident accommodation. The relatively small size of the reactor allows factory fabrication. The reactor shutdown system, inherent reactivity control and the passive decay heat removal systems result in extremely high scram and shutdown reliability. The robust primary boundary and lower internal structures allow accommodation of a hypothetical core disruptive accident and retention of a whole core molten fuel-clad mixture within the reactor. The reactor's low operating temperatures and seismic isolation result in large structural margins and high seismic capabilities. (author)

  5. Turbine inter-disk cavity cooling air compressor

    Science.gov (United States)

    Little, David Allen

    2001-01-01

    A combustion turbine may have a cooling circuit for directing a cooling medium through the combustion turbine to cool various components of the combustion turbine. This cooling circuit may include a compressor, a combustor shell and a component of the combustion turbine to be cooled. This component may be a rotating blade of the combustion turbine. A pressure changing mechanism is disposed in the combustion turbine between the component to be cooled and the combustor shell. The cooling medium preferably flows from the compressor to the combustor shell, through a cooler, the component to the cooled and the pressure changing mechanism. After flowing through the pressure changing mechanism, the cooling medium is returned to the combustor shell. The pressure changing mechanism preferably changes the pressure of the cooling medium from a pressure at which it is exhausted from the component to be cooled to approximately that of the combustor shell.

  6. Cooling system for a prestressed concrete vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Cooling system for a prestressed concrete vessel of a nuclear reactor which is faced with a steel liner and an insulating layer, and provided with cooling tubes flown through by a coolant and fixed at the prestressed concrete vessel. The cooling tubes are placed side by side and combined into several, independently operable units; they are supplied with the coolant by distributing lines operating independently, and adjacent tubes are flown through by the flow medium in opposite direction. (orig.)

  7. Post-reactor research of fuel pins, radiated in nitrogen-cooling technological channels ETC

    International Nuclear Information System (INIS)

    The state of ceramics fuel pins after long-term (about 10 hours) tests in flowing nitrogen-cooling technological channels ETC of IVG.1 reactor was researched. Rather high radiation-chemical stability of the fuel pins under definite conditions of reactor tests was determined. (author)

  8. Calculations of two-phase flows in the liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Mathematical models used for the safety analysis of liquid metal cooled fast breeder reactors are considered. Models, taking into account sodium boiling in reactor channels (one-dimensional and many-dimensional approaches), fuel cladding melting, and movement of molten materials during loss of coolant, accidents are described

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

    International Nuclear Information System (INIS)

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

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

    NARCIS (Netherlands)

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

    2009-01-01

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

  11. Comprehensive Evaluation of Secondary Cooling System of GA Siwabessy Multipurpose Reactor

    International Nuclear Information System (INIS)

    The role of heat exchanger and cooling tower in the reactor become important, because the failure of the secondary cooling system influence the reactor operation performance. The reactor has been operating 17,000 hours over, since the operation first time in 1987 years. The performance of cooling system shows decreasing and some components such as fan, drift eliminator and honeycomb of cooling tower have to be exchanged. Replacement the components must be followed with analysis of the performance of secondary cooling system comprehensively, involving to compare heat exchanger coefficient and fouling factor for heat exchanger and effectively of cooling tower by Tchebyecheff's method. The result of heat exchanger performance with the new components as follows. The heat transfer coefficient from first heat exchanger is change from 2409.9 W/m2.oC to 3512.4 W/m2.oC and second heat exchanger from 2854.9 W/m2.oC to 3150.0 W/m2.oC. In the commissioning period the values are 3720.6 W/m2.oC and 4111.9 W/m2.oC. At the other side the first line cooling tower operation capability is increased from 51.9 % to 80.3 % and second line cooling tower from 59.6 % to 92.5 %. The performance of secondary cooling after components have been replaced and repaired meet the criteria of acceptable operation condition. (author)

  12. Optimization of air ducts for nuclear reactor power generation station

    International Nuclear Information System (INIS)

    In the optimization study on the heating, ventilating and air conditions system in Nuclear Reactor Power Generation Station, proper arrangement of air ducts has been studied using the experimental and analytical investigation from a viewpoint of duct arrangement optimization. This study consists of two parts. Part I is optimization of air ducts in the corridors and Part II is optimization of air duct in each room. In part I, from viewpoints of confinement of radioactive materials in facilities having possible radioactive contamination and improvement of thermal environment for workers, the authors have studied air ducts system in which fresh air is supplied to corridors and heat removal and ventilation for each room are performed by transferring air from the corridors, instead of current ducts system with supply duct to each room. In part II, the condenser room with complex configuration and large space, and the electrical equipment room with simple space are selected for model areas. Based on these studies, experimental and analytical investigation (using a three-dimensional thermal hydraulic analysis) technique has been established, and the effective design method for duct arrangement of HVAC design has been verified for Boiling Water Reactor Power Station. The air-duct arrangements optimized in this study are applied to an Advanced Boiling Water Reactor Power Station in trial and reduction of the air-duct quantity is confirmed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

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

    International Nuclear Information System (INIS)

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

  16. Experimental investigation of an innovative low electricity consumption air cooling system for small building application

    Energy Technology Data Exchange (ETDEWEB)

    Bakos, G.C.; Tsagas, N.F. [Democritus University of Thrace, Department of Electrical and Computer Engineering, Energy Economics Laboratory, 67 100 Xanthi (Greece)

    2009-10-15

    This paper investigates the development of a low cost air cooling system used for small scale applications. Its operation is based on the rational and efficient use of energy available in conventional freezers, used for other purposes than air conditioning, to cool small areas. In this paper, the proposed air cooling system is applied to small kiosk prototype (called ''periptero''), found very often in the streets of Greek cities and villages, used to sell goods to people. The proposed air cooling system is using the low temperature air inside the freezers to cool the air inside the kiosk and provide a better living environment for the owner, particularly during summer period. The basic principle of the system operation and the experimental installation will be described and the experimental results will be presented and discussed. (author)

  17. Influence of BladeCooling on the Efficiency of Humid Air Turbine

    Directory of Open Access Journals (Sweden)

    Janusz Skorek

    2000-03-01

    Full Text Available

    The influence of an open system blade cooling on the efficiency of the humid air turbine has been analyzed. It has been assumed that three stages of the turbine are cooled by means of compressed air taken from the outlet of the compressor. The thermodynamic analysis showed, that the deleterious impact of blade cooling results mainly from the irreversibility of mixing of cooling air with the working fluid. The influence of the blade cooling on the optimum operational parameters of HAT turbine (compression ratio, excess air, air humidity and preheated air temperature has been analyzed too.

    • This paper was presented at the ECOS’99 Conference in Tokyo, June 8-10, 1999 

  18. Experimental studies on heat transfer in external cooling of the reactor pressure vessel

    International Nuclear Information System (INIS)

    The filling of the reactor cavity by accidental initiation of the containment spray system, while reactor is in full power, could have severe consequences. If the relatively cold water suddenly cools down the wall of fully pressurized reactor pressure vessel and a crack is assumed to be located on the outer surface of the vessel, the induced thermal stresses might damage the pressure vessel wall. The effects of the inadvertent cooling and pressurized thermal shock (PTS) were studied experimentally at Lappeenranta University of Technology and the heat transfer coefficients gained from the experimental results were compared with calculations. (author)

  19. Retrofitting the instrumentation and control system of primary cooling circuit from TRIGA INR 14 MW reactor

    International Nuclear Information System (INIS)

    Activities of retrofitting the instrumentation and control system from TRIGA INR primary cooling circuit consists in replacement of actual system for: - parameter measurement; - safety; - reactor external scramming; - protection, command and supply for electrical elements of the system. This retrofitting project is designed to ensure the necessary features of reactor external safety and for technological parameter measurement. The new safety system of main cooling circuit is completely separated from its operating system and is arranged in a panel assembly in reactor control room. The operating system has the following features: - data acquisition; - parameter value and state of command elements displaying; - command elements on hierarchical levels; - operator information through visual and acoustic alarm. (authors)

  20. Studies of a self-cooled lead lithium blanket for HiPER reactor

    International Nuclear Information System (INIS)

    Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall - Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates. (authors)

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

    International Nuclear Information System (INIS)

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

  2. Emergency cooling down of fast-neutron reactors by natural convection (a review)

    Science.gov (United States)

    Zhukov, A. V.; Sorokin, A. P.; Kuzina, Yu. A.

    2013-05-01

    Various methods for emergency cooling down of fast-neutron reactors by natural convection are discussed. The effectiveness of using natural convection for these purposes is demonstrated. The operating principles of different passive decay heat removal systems intended for cooling down a reactor are explained. Experimental investigations carried out in Russia for substantiating the removal of heat in cooling down fast-neutron reactors are described. These investigations include experimental works on studying thermal hydraulics in small-scale simulation facilities containing the characteristic components of a reactor (reactor core elements, above-core structure, immersed and intermediate heat exchangers, pumps, etc.). It is pointed out that a system that uses leaks of coolant between fuel assemblies holds promise for fast-neutron reactor cooldown purposes. Foreign investigations on this problem area are considered with making special emphasis on the RAMONA and NEPTUN water models. A conclusion is drawn about the possibility of using natural convection as the main method for passively removing heat in cooling down fast-neutron reactors, which is confirmed experimentally both in Russia and abroad.

  3. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  4. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    International Nuclear Information System (INIS)

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  5. Dynamic stability boundaries of a liquid metal cooled reactor system

    International Nuclear Information System (INIS)

    Part of the reactor design process is the assessment of the impact of different design changes on predefined performance criteria including stability of the reactor system under different conditions. This work focuses on the stability analysis of a combined reactor and primary heat transport system where system parameters are free to vary, with particular interest in low reactor power, low reactor coolant flow conditions. Such conditions might be encountered, for example, after a loss-of-flow without scram in some passively safe reactor designs. Linear stability analysis based methods are developed to find the stability regions, stability boundary surface in system parameter space, and frequency of oscillation at oscillatory instability boundaries. Models are presented for the reactor, detailed thermal hydraulic reactivity feedback associated with coolant outlet and inlet temperatures, decay heat, and primary system. Developed stability analysis tools are applied to the system model. System parameters include integral reactivity parameters, decay heat primary system mass, coolant flow, and natural circulation flow. The resulting stability boundary surface and its associated frequency of oscillation surface in multidimensional system parameter space show the effect of system parameter changes. By adopting model parameters for a reactor design, a stability prediction procedure is illustrated

  6. Proceedings of the workshop on the cooling of advanced reactors

    International Nuclear Information System (INIS)

    Nuclear power has become to meet electric power demand by considerable proportion, and the peaceful utilization of atomic energy steadily returns scientific and technological results to the society. As to the problem of 'Heat removal from high performance nuclear reactors' taken up successively since the last year, there are the problem of heat transport in the reactors of new types as the source of energy supply, especially the pursuit from the viewpoint of the improvement of safety and reliability related to thermal engineering, and regarding nuclear reactors, the problem of the design and operation control of experimental facilities under the utmost condition in the core and its vicinity, not only the problem of reactors proper. Particularly regarding research reactors, precision has become to be demanded in addition to the safety and reliability thermally for various facilities. In the workshop of this year, the presenting of reports and discussion were carried out from the standpoint of thermal engineering on fast reactors and light water reactors of next generation, new research reactors and experimental facilities. (K.I.)

  7. Operating Experience with the BN-600 Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The report considers the main design features of the BN-600 liquid metal fast reactor. The performance indicators achieved for 32 years of operation are given. The measures taken to enhance BN-600 reactor power unit safety and replace and extend its equipment lifetime and their results allowed the design lifetime of the power unit to be extended up to 40 years (until 31 March 2020) are presented. The considered integrated material, methodological and theoretical investigations justifying the serviceability of the irreplaceable components of the BN-600 reactor facility have shown that the strength conditions have not been violated in any of the critical reactor components after 45 years of operation. The results, both of the actions taken to enhance the BN-600 reactor power unit safety and corrective measures related to the events at the Fukushima nuclear power plant, allow the safety of the power unit exposed to any possible extreme external impact to be improved. (author)

  8. 16 CFR Appendix H to Part 305 - Cooling Performance and Cost for Central Air Conditioners

    Science.gov (United States)

    2010-01-01

    ... 16 Commercial Practices 1 2010-01-01 2010-01-01 false Cooling Performance and Cost for Central Air Conditioners H Appendix H to Part 305 Commercial Practices FEDERAL TRADE COMMISSION REGULATIONS UNDER SPECIFIC... RULEâ) Pt. 305, App. H Appendix H to Part 305—Cooling Performance and Cost for Central Air...

  9. Design guidelines for the forced-air cooling process of strawberries

    Energy Technology Data Exchange (ETDEWEB)

    Ferrua, M.J.; Singh, R.P. [Department of Biological and Agricultural Engineering, University of California, Davis, CA 95616 (United States)

    2009-12-15

    The complex structure of the package systems currently used by the strawberry industry has prevented an efficient design of the forced-air cooling process. In this study, we investigated the mechanisms through which different design parameters affect the rate and uniformity of cooling, using a previously validated computational fluid dynamics model. The results indicated that the vent area has a significant effect on the cooling rate, but not on its uniformity. A design that reduces bypassing will not necessarily increase the cooling rate, because there is less bypass air to cool down the air exiting from each clamshell, so that the air entering the next clamshell becomes warmer. Periodic airflow reversal improves the rate and homogeneity of the cooling process. (author)

  10. Experimental feasibility study of radial injection cooling of three-pad radial air foil bearings

    Science.gov (United States)

    Shrestha, Suman K.

    Air foil bearings use ambient air as a lubricant allowing environment-friendly operation. When they are designed, installed, and operated properly, air foil bearings are very cost effective and reliable solution to oil-free turbomachinery. Because air is used as a lubricant, there are no mechanical contacts between the rotor and bearings and when the rotor is lifted off the bearing, near frictionless quiet operation is possible. However, due to the high speed operation, thermal management is one of the very important design factors to consider. Most widely accepted practice of the cooling method is axial cooling, which uses cooling air passing through heat exchange channels formed underneath the bearing pad. Advantage is no hardware modification to implement the axial cooling because elastic foundation structure of foil bearing serves as a heat exchange channels. Disadvantage is axial temperature gradient on the journal shaft and bearing. This work presents the experimental feasibility study of alternative cooling method using radial injection of cooling air directly on the rotor shaft. The injection speeds, number of nozzles, location of nozzles, total air flow rate are important factors determining the effectiveness of the radial injection cooling method. Effectiveness of the radial injection cooling was compared with traditional axial cooling method. A previously constructed test rig was modified to accommodate a new motor with higher torque and radial injection cooling. The radial injection cooling utilizes the direct air injection to the inlet region of air film from three locations at 120° from one another with each location having three axially separated holes. In axial cooling, a certain axial pressure gradient is applied across the bearing to induce axial cooling air through bump foil channels. For the comparison of the two methods, the same amount of cooling air flow rate was used for both axial cooling and radial injection. Cooling air flow rate was

  11. Analysis of a loss of forced cooling test using the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is the first High Temperature Gas-cooled Reactor (HTGR) built at the Oarai Research and Development Center of JAEA, with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950degC (Saito, 1994). Test researches are being conducted using the HTTR to improve HTGR technologies and to collaborate with domestic industries to contribute to foreign projects for acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are being developed using data obtained with the HTTR, which include reactor kinetics, thermal-hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). A three-gas-circulators trip test and a vessel-cooling-system stop test were planned as a loss-of-forced-cooling test and demonstrate the inherent safety features of HTGR. The vessel-cooling-system stop test consists of stopping the vessel-cooling-system located outside the reactor pressure vessel (RPV), to remove the residual heat of the reactor core as soon as the three-gas-circulators are tripped. All three-gas-circulators is tripped at 9 MW. The primary coolant flow rate is reduced from the rated 45 t/h to 0 t/h. The control rods are not inserted into the core and the reactor power control system does not operated. A core dynamics analysis of the loss-of-forced-cooling test of the HTTR is performed. Analytical results for the reactor transient during the test are presented in this report. It is determined that the reactor power immediately decreases to the decay heat level due to the negative reactivity feedback effect of the core, even though the reactor shutdown system is not operational, and that the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Furthermore, the relation between the reactivities (namely, the Doppler, moderator temperature, and

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

    International Nuclear Information System (INIS)

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

  13. Coated particle fuel for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 μm diameter UO2 kernel of 10% enrichment is surrounded by a 100 μm thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 μm thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level

  14. Evaluation of a Design Concept for the Combined Air-water Passive Cooling PAFS+

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sung Won; Kwon, Taesoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The APR+ system provides the Passive Auxiliary Feed-water System (PAFS) for the passive cooling capability. However, the current design requirement for working time for the PAFS is about 8 hours only. Thus, current working time of PAFS can not meet the required 72 hours cooling capability for the long term SBO situation. To meet the 72 hours cooling, the pool capacity should be almost 3∼4 times larger than that of current water cooling tank. In order to continue the PAFS operation for 72 hours, a new passive air-water combined cooling system is proposed. This paper provides the feasibility study on the combined passive air-water cooling system. Figure 1 and 2 show the conceptual difference of the PAFS and combined passive air-water cooling system, respectively. Simple performance evaluation of the passive air cooling heat exchanger has been conducted by the MARS calculation. For the postulated FLB scenario, 4800 heat exchanger tubes and 5 m/s air velocity are not sufficient to sustain the PCCT pool level for 72 hour cooling. Further works on the system design and performance enhancing plan are required to fulfill the 72 hours long term passive cooling.

  15. Evaluation of a Design Concept for the Combined Air-water Passive Cooling PAFS+

    International Nuclear Information System (INIS)

    The APR+ system provides the Passive Auxiliary Feed-water System (PAFS) for the passive cooling capability. However, the current design requirement for working time for the PAFS is about 8 hours only. Thus, current working time of PAFS can not meet the required 72 hours cooling capability for the long term SBO situation. To meet the 72 hours cooling, the pool capacity should be almost 3∼4 times larger than that of current water cooling tank. In order to continue the PAFS operation for 72 hours, a new passive air-water combined cooling system is proposed. This paper provides the feasibility study on the combined passive air-water cooling system. Figure 1 and 2 show the conceptual difference of the PAFS and combined passive air-water cooling system, respectively. Simple performance evaluation of the passive air cooling heat exchanger has been conducted by the MARS calculation. For the postulated FLB scenario, 4800 heat exchanger tubes and 5 m/s air velocity are not sufficient to sustain the PCCT pool level for 72 hour cooling. Further works on the system design and performance enhancing plan are required to fulfill the 72 hours long term passive cooling

  16. Data on test results of vessel cooling system of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1989-08-01

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

  18. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D.; Herring, J.S.; Macdonald, P.E. [Idaho National Engineering and Environment Lab., Advanced Nuclear Energy, Idaho (United States)

    2001-07-01

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  19. Performance Comparison of Metallic, Actinide Burning Fuel in Lead-Bismuth and Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-04-01

    Various methods have been proposed to “incinerate” or “transmutate” the current inventory of trans-uranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non-fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years.

  20. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  1. During air cool process aerosol absorption detection with photothermal interferometry

    Science.gov (United States)

    Li, Baosheng; Xu, Limei; Huang, Junling; Ma, Fei; Wang, Yicheng; Li, Zhengqiang

    2014-11-01

    This paper studies the basic principle of laser photothermal interferometry method of aerosol particles absorption coefficient. The photothermal interferometry method with higher accuracy and lower uncertainty can directly measure the absorption coefficient of atmospheric aerosols and not be affected by scattered light. With Jones matrix expression, the math expression of a special polarization interferometer is described. This paper using folded Jamin interferometer, which overcomes the influence of vibration on measuring system. Interference come from light polarization beam with two orthogonal and then combine to one beam, finally aerosol absorption induced refractive index changes can be gotten with four beam of phase orthogonal light. These kinds of styles really improve the stability of system and resolution of the system. Four-channel detections interact with interference fringes, to reduce the light intensity `zero drift' effect on the system. In the laboratory, this device typical aerosol absorption index, it shows that the result completely agrees with actual value. After heated by laser, cool process of air also show the process of aerosol absorption. This kind of instrument will be used to monitor ambient aerosol absorption and suspended particulate matter chemical component. Keywords: Aerosol absorption coefficient; Photothermal interferometry; Suspended particulate matter.

  2. Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

    International Nuclear Information System (INIS)

    The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

  3. General features of direct-cycle, supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The concept of direct-cycle, supercritical-pressure, light-water-cooled reactors is developed. Breeding is possible in the tight lattice core. The power output can be maximized in the fast converter reactor. The gross thermal efficiency of the high temperature reactor adopting Inconel as fuel cladding is expected to be 44.8%. The plant system is similar to the supercritical-fossil-fired power plant which adopts once-through type coolant circulation system. The volume and height of the containment are approximately half of the BWR. The basic safety principles follows those of LWRs. The reactor will solve the economic problems of LWR and LMFBR

  4. COMMISION DEBUGGING OF AIR-COOLED ISLAND SYSTEMS FOR 600 MW AIR-COOLED UNITS%600 MW空冷机组空冷岛的调试

    Institute of Scientific and Technical Information of China (English)

    杨海生; 李路江; 吴瑞涛; 刘春报; 刘红霞

    2009-01-01

    The problems appeared in commission debugging of the air-cooled island systems for two 600 MW air-cooled units of Guodian Longshan Power Generation Co Ltd,such as undue fast vacuum drop in air-cooled island during start-up of said units after their shutdown in winter, the rapid rise of back-pressure in said air-cooled isrand due to full-load operation of all cooling air fans and air-leakage in the steam seal system, etc. , have been analysed, and corresponding preventive measures a-dopted for above-mentioned problems in the commission debugging process being given. Regarding to a part of problems, which hadn't been solved in debugging, some concrete recommendations have been put forward.%对国电河北龙山发电有限公司2×600 MW空冷机组空冷岛的调试中出现的问题进行了分析,如冬季停机后起动空冷岛真空下降过快,全部空冷风机全负荷运转,汽封系统漏空气空冷岛背压急剧上升等,给出了调试中对上述问题采取的相关防范措施,并对调试中部分未能解决的问题,提出了具体建议.

  5. Simulation and Optimization of Air-Cooled PEMFC Stack for Lightweight Hybrid Vehicle Application

    Directory of Open Access Journals (Sweden)

    Jingming Liang

    2015-01-01

    Full Text Available A model of 2 kW air-cooled proton exchange membrane fuel cell (PEMFC stack has been built based upon the application of lightweight hybrid vehicle after analyzing the characteristics of heat transfer of the air-cooled stack. Different dissipating models of the air-cooled stack have been simulated and an optimal simulation model for air-cooled stack called convection heat transfer (CHT model has been figured out by applying the computational fluid dynamics (CFD software, based on which, the structure of the air-cooled stack has been optimized by adding irregular cooling fins at the end of the stack. According to the simulation result, the temperature of the stack has been equally distributed, reducing the cooling density and saving energy. Finally, the 2 kW hydrogen-air air-cooled PEMFC stack is manufactured and tested by comparing the simulation data which is to find out its operating regulations in order to further optimize its structure.

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

    International Nuclear Information System (INIS)

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R and D) that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. This paper presents current R and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs

  7. Thermal hydraulics of the very high temperature gas cooled reactor

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-01

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

  9. Shutdown cooling helium circulator design considerations for MHTGR [Modular High Temperature Gas-Cooled Reactor] power plant

    International Nuclear Information System (INIS)

    The Modular High Temperature Gas-Cooled Reactor (MHTGR) plant embodies a shutdown cooling system to expedite plant cooldown for refueling, maintenance, and repair in the event that the main cooling loop is unavailable. This is a non safety related system. A key component in this system, is a helium circulator. Oriented vertically, the rotating assembly in this machine is supported on active magnetic bearings, and the radial flow compressor is driven by a submerged induction electric motor rated at 160 kW(e). This paper gives details of the circulator design considerations and includes topics related to the machine operation and maintenance, and the technology base. 12 refs., 11 figs., 3 tabs

  10. The New Emergency Core Cooling (NECC) system for the National Research Universal (NRU) reactor

    International Nuclear Information System (INIS)

    The New Emergency Core Cooling (NECC) system is the penultimate of seven major safety upgrades being implemented at the National Research Universal (NRU) Reactor in Chalk River. The NECC upgrade was designed to improve the original systems for core cooling in the event of an unisolable failure within the primary cooling circuit. The NECC upgrade ensures that water is automatically made available to the emergency cooling circuit pumps in the event of a break. Reactor core cooling is achieved from the discharge of these pumps which distribute emergency coolant to the individual fuel rods. Heated water from the vessel returns to the heat exchangers within the emergency cooling circuits for heat removal to the secondary coolant. The NECC upgrade significantly improves protection for a wide range of Loss Of Coolant Accidents (LOCAs) through the use of design features such as component redundancy, automatic initiation and hazard qualification. The introduction of the NECC upgrade combined with previous improvements in liquid confinement capability provide a closed loop system that ensures stable long term reactor core cooling. CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) analysis was performed to assess the NECC upgrade and to validate the design for credible leak scenarios. (author)

  11. Conceptual design of reactor TRIGA PUSPATI (RTP) spent fuel pool cooling system

    International Nuclear Information System (INIS)

    After undergo about 30 years of safe operation, Reactor TRIGA PUSPATI (RTP) was planned to be upgraded to ensure continuous operation at optimum safety condition. In the meantime, upgrading is essential to get higher flux to diversify the reactor utilization. Spent fuel pool is needed for temporary storage of the irradiated fuel before sending it back to original country for reprocessing, reuse after the upgrading accomplished or final disposal. The irradiated fuel elements need to be secure physically with continuous cooling to ensure the safety of the fuels itself. The decay heat probably still exist even though the fuel elements not in the reactor core. Therefore, appropriate cooling is required to remove the heat produced by decay of the fission product in the irradiated fuel element. The design of spent fuel pool cooling system (SFPCS) was come to mind in order to provide the sufficient cooling to the irradiated fuel elements and also as a shielding. The spent fuel pool cooling system generally equipped with pumps, heat exchanger, water storage tank, valve and piping. The design of the system is based on criteria of the primary cooling system. This paper provides the conceptual design of the spent fuel cooling system. (author)

  12. Control of cooling processes with forced-air aimed at efficiency energetic and the cooling time of horticultural products

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Joao Carlos Teles Ribeiro da; Mederos, Barbara Janet Teruel [Universidade Estadual de Campinas (FEAGRI/UNICAMP), SP (Brazil). Fac. de Engenharia Agricola

    2008-07-01

    The application of cooling technologies for the conservation of horticultural products is one of the stages the Cold Chain. In Brazil particularly, as a country with tropical climate with average high temperature almost all year, the application of these technologies is very important because the shelf-life of fresh horticultural products, with quality that the market demands, is directly related to temperature. In particular, the systems of forced air cooling operate according to the flow of air predetermined in the project according to the quantity of product to cool. When actual conditions differ from considerations of the project, as to the quantity of product, a situation very common in agricultural properties and packing houses, the fan will continue providing the nominal flow rate, causing alteration of the cost-benefit relation of process. This project aims at the development of a micro-processing equipment (output current of 4 to 20 mA) to control the rotational speed of the motor of the fan systems, air forced through an inverter of frequency. The objective is development of a Man-Machine Interface, based on an algorithm, which, through the introduction of mass product data and the automatic acquisition of data from temperature of the product and the camera, is calculated the cooling time. The rotation of the engine fan will be amended automatically, to maintain air flow with a proper cost-benefit, in connection with the reduction of cooling time, energy consumption, for the increasing the shelf life of products. (author)

  13. Air Corrosivity in U.S. Outdoor-Air-Cooled Data Centers is Similar to That in Conventional Data Centers

    Energy Technology Data Exchange (ETDEWEB)

    Coles, Henry C.; Han, Taewon; Price, Phillip N.; Gadgil, Ashok J.; Tschudi, William F.

    2011-07-17

    There is a concern that environmental-contamination caused corrosion may negatively affect Information Technology (IT) equipment reliability. Nineteen data centers in the United States and two in India were evaluated using Corrosion Classification Coupons (CCC) to assess environmental air quality as it may relate IT equipment reliability. The data centers were of two basic types: closed and outside-air cooled. A closed data center provides cool air to the IT equipment using air conditioning in which only a small percent age of the recirculation air is make-up air continuously supplied from outside to meet human health requirements. An outside-air cooled data center uses outside air directly as the primary source for IT equipment cooling. Corrosion measuring coupons containing copper and silver metal strips were placed in both closed and outside-air cooled data centers. The coupons were placed at each data center (closed and outside-air cooled types) with the location categorized into three groups: (1) Outside - coupons sheltered, located near or at the supply air inlet, but located before any filtering, (2) Supply - starting just after initial air filtering continuing inside the plenums and ducts feeding the data center rooms, and (3) Inside located inside the data center rooms near the IT equipment. Each coupon was exposed for thirty days and then sent to a laboratory for a corrosion rate measurement analysis. The goal of this research was to investigate whether gaseous contamination is a concern for U.S. data center operators as it relates to the reliability of IT equipment. More specifically, should there be an increased concern if outside air for IT equipment cooling is used To begin to answer this question limited exploratory measurements of corrosion rates in operating data centers in various locations were undertaken. This study sought to answer the following questions: (1) What is the precision of the measurements (2) What are the approximate statistical

  14. Characteristic behavior of pebble-bed modular high-temperature gas-cooled reactor during loss of forced cooling accidents

    International Nuclear Information System (INIS)

    Based on the preliminary design of the Pebble-bed Modular High-Temperature Gas-cooled Reactor(HTR-PM), two cases of loss of forced cooling accident (DLOFC and PLOFC) were studied by the help of the software THERMIX. The key parameters including reactor power, temperature distributions of the core and pressure vessel, and the decay power removal by the passive residual heat remove system(RHRS) were compared in detail. Some parameter uncertainties were analyzed in order to evaluate the safety margin of the maximal fuel temperature during LOFC. The calculated results show that, the decay heat in the LOFC accidents can be removed from the reactor core solely by means of physical processes in a passive way, so that the temperature limits of fuel and components are still obeyed, which can effectively keep the integrality of the fuel particles to avoid massive fission products release. It also illustrates that the HTR-PM can reach 250 MW reactor power per unit and still can keep the inherent safety, which will be helpful to the further detail design of the HTR-PM demonstrating power plant project. (authors)

  15. Citywide Impacts of Cool Roof and Rooftop Solar Photovoltaic Deployment on Near-Surface Air Temperature and Cooling Energy Demand

    Science.gov (United States)

    Salamanca, F.; Georgescu, M.; Mahalov, A.; Moustaoui, M.; Martilli, A.

    2016-04-01

    Assessment of mitigation strategies that combat global warming, urban heat islands (UHIs), and urban energy demand can be crucial for urban planners and energy providers, especially for hot, semi-arid urban environments where summertime cooling demands are excessive. Within this context, summertime regional impacts of cool roof and rooftop solar photovoltaic deployment on near-surface air temperature and cooling energy demand are examined for the two major USA cities of Arizona: Phoenix and Tucson. A detailed physics-based parametrization of solar photovoltaic panels is developed and implemented in a multilayer building energy model that is fully coupled to the Weather Research and Forecasting mesoscale numerical model. We conduct a suite of sensitivity experiments (with different coverage rates of cool roof and rooftop solar photovoltaic deployment) for a 10-day clear-sky extreme heat period over the Phoenix and Tucson metropolitan areas at high spatial resolution (1-km horizontal grid spacing). Results show that deployment of cool roofs and rooftop solar photovoltaic panels reduce near-surface air temperature across the diurnal cycle and decrease daily citywide cooling energy demand. During the day, cool roofs are more effective at cooling than rooftop solar photovoltaic systems, but during the night, solar panels are more efficient at reducing the UHI effect. For the maximum coverage rate deployment, cool roofs reduced daily citywide cooling energy demand by 13-14 %, while rooftop solar photovoltaic panels by 8-11 % (without considering the additional savings derived from their electricity production). The results presented here demonstrate that deployment of both roofing technologies have multiple benefits for the urban environment, while solar photovoltaic panels add additional value because they reduce the dependence on fossil fuel consumption for electricity generation.

  16. Cooling Performance Characteristics on Mobile Air-Conditioning System for Hybrid Electric Vehicles

    OpenAIRE

    Ho-Seong Lee; Moo-Yeon Lee

    2013-01-01

    This study investigates the cooling performance characteristics of the mobile air-conditioning system using R744 (CO2) for the hybrid electric vehicle as an alternative to both the R-134a and the conventional air-conditioning system. The developed air-conditioning system is operated with an electric driven compressor in the battery driving mode and a belt driven compressor in the engine driving mode. The cooling performance characteristics of the developed system have been analyzed by experim...

  17. Optimum design of bipolar plates for separate air flow cooling system of PEM fuel cells stacks

    Science.gov (United States)

    Franco, Alessandro

    2015-12-01

    The paper discusses about thermal management of PEM fuel cells. The objective is to define criteria and guidelines for the design of the air flow cooling system of fuel cells stacks for different combination of power density, bipolar plates material, air flow rate, operating temperature It is shown that the optimization of the geometry of the channel permits interesting margins for maintaining the use of separate air flow cooling systems for high power density PEM fuel cells.

  18. Benchmark problem for International Atomic Energy Agency (IAEA) coordinated research program (CRP) on gas-cooled reactor (GCR) afterheat Removal

    International Nuclear Information System (INIS)

    In IAEA CRP on 'Heat Transport and Afterheat Removal for GCRs under Accident Conditions', experimental data of the JAERI's cooling panel test apparatus were selected as benchmark problems to verify the validity of computational codes for design and evaluation of the performance of heat transfer and temperature distribution of components in the cooling panel system of the HTGR. The test apparatus was composed of a pressure vessel (P.V) with 1m in diameter and 3m in height, containing heaters with the maximum heating rate of 100kW simulating decay heat, cooling panels surrounding the P.V and the reactor cavity occupied by air at the atmospheric pressure. Seven experimental data were established as benchmark problems to evaluate the effect of natural convection of superheated gas on temperature distribution of the P.V and the performance of heat transfer of both the water and the air cooling panel systems. The analytical code THANPACST2 was applied to analyze two benchmark problems to verify the validity of the analytical methods and models proposed. Under the conditions at helium gas pressure of 0.73MPa and temperature of 210degC in the P.V of the water cooling panel system, temperatures of the P.V were well estimated within the errors of -14% to +27% compared with the experimental data. The analyses indicated that the heat transferred to the cooling panel was 11.4% less than the experimental value and the heat transferred by thermal radiation was 74.4% of the total heat input. (author)

  19. Impact of the use of a hybrid turbine inlet air cooling system in arid climates

    International Nuclear Information System (INIS)

    Graphical abstract: Cooling the air entering the compressor section of a gas turbine is a proven method of increasing turbine power output, especially during peak summer demand, and it is increasingly being used in powerplants worldwide. Two turbine inlet air cooling (TIAC) systems are widely used: evaporative cooling and mechanical chilling. In this work, the prospects of using a hybrid turbine inlet air cooling (TIAC) system are investigated. The hybrid system consists of mechanical chilling followed by evaporative cooling. Such a system is capable of achieving a significant reduction in inlet air temperature that satisfies desired power output levels, while consuming less power than conventional mechanical chilling and less water than conventional evaporative cooling, thus combining the benefits of both approaches. Two hybrid system configurations are studied. In the first configuration, the first stage of the system uses water-cooled chillers that are coupled with dry coolers such that the condenser cooling water remains in a closed loop. In the second configuration, the first stage of the system uses water-cooled chillers but with conventional cooling towers. An assessment of the performance and economics of those two configurations is made by comparing them to conventional mechanical chilling and using realistic data. It was found that the TIAC systems are capable of boosting the power output of the gas turbine by 10% or more (of the power output of the ISO conditions). The cost operation analysis shows clearly the hybrid TIAC method with wet cooling has the advantage over the other methods and It would be profitable to install it in the new gas turbine power plants. The figure below shows a comparison of the water consumption for the three different cases. - Highlights: • New hybrid system for the turbine inlet air cooling is studied. • Hybrid system of mechanical chilling followed by evaporative cooling is used. • Hybrid turbine inlet air cooling

  20. Towards the Characterization of the Bubble Presence in Liquid Sodium of Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    In a Sodium cooled Fast Reactors (SFR), different phenomena such as gas entrainment or nucleation can lead to gaseous micro-bubbles presence in the liquid sodium of the primary vessel. Although this free gas presence has no direct impact on the core neutronics, the French Atomic Energy and Alternative Energies Commission (CEA) currently works on its characterization to, among others, check the absence of risk of large gas pocket formation and to assess the induced modifications of the sodium acoustic properties. The main objective is to evaluate the void fraction values (volume fraction of free gas) and the radii histogram of the bubbles present in liquid sodium. Acoustics and electromagnetic techniques are currently developed at CEA: - The low-frequency speed of sound measurement, which allows us to link - thanks to Wood's model - the measured speed of sound to the actual void fraction. - The nonlinear mixing of two frequencies, based on the nonlinear resonance behavior of a bubble. This technique allows knowing the radius histogram associated to a bubble cloud. Two different mixing techniques are presented in this paper: the mixing of two high frequencies and the mixing of a high and a low frequency. - The Eddy-current flowmeter (ECFM), the output signal of which is perturbed by free gas presence and in consequence allows detecting bubbles. For each technique, initial results are presented. Some of them are really promising. So far, acoustic experiments have been led with an air-water experimental set-up. Micro-bubbles clouds are generated with a dissolved air flotation device and monitored by an optical device which provides reference measurements. Generated bubbles have radii range from few micrometers to several tens of micrometers. Present and future air/water experiments are presented. Furthermore, a development plan of in-sodium tests is presented in terms of a device set-up, instrumentation, modeling tools and experiments. (authors)

  1. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 oC which is substantially lower than ∼627 oC as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

  2. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hung, T.C., E-mail: tchung@ntut.edu.t [Department of Mechanical Engineering, National Taipei University of Technology, 1, Sec. 3, Chung-hsiao E. Rd., Taipei 10608, Taiwan (China); Dhir, V.K. [Department of Mechanical and Aerospace Engineering, UCLA, CA (United States); Chang, J.C. [Graduate Institute of Mechanical and Electrical Engineering, National Taipei University of Technology, Taiwan (China); Wang, S.K. [Department of Mechanical and Automation Engineering, I-Shou University, Taiwan (China)

    2011-01-15

    Research highlights: The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW) The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to {approx}551 {sup o}C which is substantially lower than {approx}627 {sup o}C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity

  3. An evolutionary approach to advanced water cooled reactors

    International Nuclear Information System (INIS)

    Based on the result of the Feasibility Study undertaken since 1991, Indonesia may enter in the new nuclear era by introduction of several Nuclear Power Plants in our energy supply system. Requirements for the future NPP's are developed in two step approach. First step is for the immediate future that is the next 50 years where the system will be dominated by A-LWR's/A-PHWR's and the second step is for the time period beyond 50 years in which new reactor systems may start to dominate. The integral reactor concept provides a revolutionary improvements in terms of conceptual and safety. However, it creates a new set of complex machinery and operational problems of its own. The paper concerns with a brief description of nuclear technology status in Indonesia and a qualitative assessment of integral reactor concept. (author)

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

    International Nuclear Information System (INIS)

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

  5. Probabilistic safety assessment method for sodium fire of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    The sodium fire is a typical and distinctive hazard in sodium cooled fast reactor, which is probably one of the main contributors to the total reactor risks. In this paper, the methodology of fast reactor sodium fire risk assessment was studied, following the introduction of the sodium fire. The application of this technology in China Experimental Fast Reactor was explored, and the results show that the core damage frequency induced by the sodium fire in reactor hall is 1.19 × 10-8(reactor · year). After that, several key problems which need to be further researched in the future during the process of sodium fire probabilistic safety assessment were discussed. (authors)

  6. Uncertainties treatment in the water-cooled nuclear research reactor-thermal design and analysis

    International Nuclear Information System (INIS)

    This paper describes methods of uncertainties and its calculationprocedures for the water-cooled nuclear research reactor (i. e. WWR-M2) with a 10 MWth, in the thermal design and analysis, where the uncertainties are due to the reactor fuel coolant channel design fabrication defects (fuel meat and clad thickness uncertainties). The results show: (1) The effect of fuel meat and cladding thickness may have great influence on the distribution of the axial temperatures (cladding surface, and fuel centerline) and other parameters in the reactor core and more intense in the reactor thermal design, (2) The selection of new reactor core operating conditions and parameters due to the fuel coolant channel fabrication defects, and (3) Calculation values of the hot spot and temperatures of the WWR-M2 reactor by using different methods.(author)

  7. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  8. Status and Future Challenges of CFD for Liquid Metal Cooled Reactors

    International Nuclear Information System (INIS)

    Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Typically, sodium and lead(-alloys) are envisaged as coolants for such reactors. Obviously, in the development of these reactors, thermal-hydraulics is recognized as a key (safety) challenge. A relatively new technique to deal with thermal-hydraulics issues is Computational Fluid Dynamics (CFD). This technique is used increasingly nowadays for design and safety evaluation purposes. This paper will discuss the development status of CFD application to liquid metal cooled reactors. In addition, the main challenges for future developments will be indicated. Firstly, the technological challenges will be discussed which ask for CFD application. Afterwards, the needs for CFD development and/or validation will be discussed. The discussion will also include the need for accompanying experiments. (author)

  9. Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

    International Nuclear Information System (INIS)

    In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

  10. Verification calculations as per CFD FLOWVISION code for sodium-cooled reactor plants

    International Nuclear Information System (INIS)

    The paper studies the experience in application of CFD FlowVision software for analytical validation of sodium-cooled fast reactor structure components and the results of performed verification, namely: – development and implementation of new model of turbulent heat transfer in liquid sodium (LMS) in FlowVision software and model verification based on thermohydraulic characteristics studied by experiment at TEFLU test facility; – simulation of flowing and mixing of coolant with different temperatures in the upper mixing chamber of fast neutron reactor through the example of BN-600 (comparison with the results obtained at the operating reactor). Based on the analysis of the results obtained, the efficiency of CFD codes application for the considered problems is shown, and the proposals for CFD codes verification development as applied to the advanced sodium-cooled fast reactor designs are stated. (author)

  11. Preliminary Investigation of an Optimally Scramming Control Rod for Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    A passively safe control rod for gas-cooled reactors is proposed. This Optimally Scramming Control Rod (OSCR) is lifted out of the core region by the core coolant and descends back into the core when the coolant flow is not sufficient for core cooling purposes or in the event of depressurization. It is shown that for the current design of the OSCR, the reactor can be operated under normal lower power conditions down to about 80% of total power. It is also shown that cold shutdown can be achieved with rods of sufficiently low mass to allow naturally passive operation of the concept. (authors)

  12. A project definition for demonstrating a modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    This paper describes the modular high temperature gas-cooled reactor (MHTGR) design developed within the U.S. HTGR Program and the MHTGR demonstration project recently defined by Gas-Cooled Reactor Associates (GCRA) and its utility members. A Project Definition Study was funded by GCRA and the Tennessee Valley Authority (TVA) and was cost-shared by the participating contractors. The study considered a repowering option at an existing utility site and a stand-alone, full-plant option at a remote, government site. Through the proposed demonstration project, the MHTGR would become a commercial alternative for the U.S. utility industry

  13. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Orvis, D.D.; Raabe, P.H.

    1980-01-01

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs.

  14. Methods and technologies for cost reduction in the design of water cooled reactor power plants

    International Nuclear Information System (INIS)

    The Specialists Meeting was organized in the framework of the IAEA International Working Group on Advanced Technologies for Water-Cooled Reactors. Its purpose was to provide an international forum for review and discussion on recent results in research and development on different methods and technologies of current and advanced water-cooled reactor power plants, which can lead to reduced investment and operation, maintenance and fuel-cycle costs of the plants. 27 specialists representing 10 countries and the IAEA took part in the meeting. 10 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  15. Licensing topical report: interpretation of general design criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    This Licensing Topical Report presents a set of General Design Criteria (GDC) which is proposed for applicability to licensing of graphite-moderated, high-temperature gas-cooled reactors (HTGRs). Modifications as necessary to reflect HTGR characteristics and design practices have been made to the GDC derived for applicability to light-water-cooled reactors and presented in Appendix A of Part 50, Title 10, Code of Federal Regulations, including the Introduction, Definitions, and Criteria. It is concluded that the proposed set of GDC affords a better basis for design and licensing of HTGRs

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

    International Nuclear Information System (INIS)

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

  17. DESIGN OPTIMIZATION OF A FUSION REACTOR DIVERTOR COOLING FINGER DESIGN FOR DECREASED THERMO-MECHANICAL STRESSES

    OpenAIRE

    MATTEOLI CAMILLA; Martin, Oliver; SIMONOVSKI IGOR

    2012-01-01

    In this paper a design of a divertor cooling finger for a fusion reactor is looked at with the aim of reducing the thermo-mechanical stresses. The function of a divertor in a fusion reactor is to reduce the dilution of the plasma by removing alpha particles, helium and other impurities. In addition, this component has to remove approximately 15% of the total thermal power. The divertor is therefore exposed to a significant thermal load, and during the operation it has to be actively cooled, w...

  18. Improving economics and safety of water cooled reactors. Proven means and new approaches

    International Nuclear Information System (INIS)

    Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

  19. Objective Provision Tree (OPT) in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Application to the safety function of residual heat removal As part of the project Safety Assessment for Reactor of GEN-IV (SARGEN IV) has been implemented the methodology ISAM from the IAEA to the safety assessment of new sodium reactor designs. Within the ISAM, a new tool to facilitate this assessment is the Objective Provision Tree (OPT) which documents the provisions necessary for each of the levels of defense in depth, as well as for each critical function of security. Due to the design innovations that have sodium reactors, the evaluation of safety and licensing of these reactors requires special considerations. In this work we have analyzed the mechanisms of failure of the safety function concerning the evacuation of waste heat, and have been proposed different provisions for each of the first three levels of defense in depth. The main result of this work is reflected in the elaboration of the OPTs, one for each of the first three levels of defense in depth for the safety of evacuation of residual heat function. These trees represent in a schematic way the provisions necessary to comply with the objectives of each level which are respectively: 1) deviations from normal operation, 2) control of abnormal operation and fault detection and 3) incidental control.

  20. Improvement of the Decay Heat Removal Characteristics of the Generation IV Gas-cooled Fast Reactor

    OpenAIRE

    Epiney, Aaron Simon

    2010-01-01

    Gas cooling in nuclear power plants (NPPs) has a long history, the corresponding reactor types developed in France, the UK and the US having been thermal neutron-spectrum systems using graphite as the moderator. The majority of NPPs worldwide, however, are currently light water reactors, using ordinary water as both coolant and moderator. These NPPs – of the so-called second generation – will soon need replacement, and a third generation is now being ...

  1. IAEA Workshop (Training Course) on Codes and Standards for Sodium Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    The training course consisted of lectures and Q&A sessions. The lectures dealt with the history of the development of Design Codes and Standards for Sodium Cooled Fast Reactors (SFRs) in the respective country, the detailed description of the current design Codes and Standards for SFRs and their application to ongoing Fast Reactor design projects, as well as the ongoing development work and plans for the future in this area. Annex 1 contains the detailed Workshop program

  2. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    OpenAIRE

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable future high-tech process heat applications with minimized R&D. The development of HTR started in the 1950s as a graphite moderated and helium cooled reactor. This concept featured important inherent...

  3. The key device--elevator in 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The basic structure, working principle and behavior of the control system of the elevator in 10 MW high temperature gas-cooled reactor (HTR-10) are researched. The five-phase hybrid stepping motor and the closed-loop control are adopted in the construction design of the elevator. About 20000 fuel elements and graphite balls were transported into the reactor core by the elevator to achieve the critical loading for HTR-10

  4. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.)

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

    International Nuclear Information System (INIS)

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

  6. Three-dimensional imaging and precision metrology for liquid-salt-cooled reactors

    International Nuclear Information System (INIS)

    The liquid-salt-cooled very high temperature reactor, also called the Advanced High-Temperature Reactor (AHTR), is a new large high-temperature reactor concept that combines in a novel way four established technologies: (1) coated-particle graphite-matrix nuclear fuels, (2) Brayton power cycles, (3) passive safety systems and plant designs previously developed for liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants. The AHTR will require refueling, in-service inspection, and maintenance (RIM) with supporting instrumentation systems. The fluoride salts that are being evaluated as potential reactor coolants have melting points between 350 and 500 deg. C, values that imply minimum RIM temperatures between 400 and 550 deg. C. These salts are transparent over a wider range of the light spectrum than is water. The high temperatures, the optical characteristics of the coolant, and advances in metrology may enable the use of lasers to create three-dimensional images of the reactor interior to assist refueling, monitor vibrations in components, map fluid flow, and enable inspections of internal reactor components. A description of the reactor and an initial evaluation of the use of optical techniques for AHTR instrumentation are provided. (authors)

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

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

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

  8. Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor

    International Nuclear Information System (INIS)

    The paper describes elements of design consideration of supercritical-pressure, light water cooled reactors as well as the status and prospects of the research and development. It summarizes the results of the conceptual design study at the University of Tokyo from 1989. The research and development started in Japan, Europe and USA. The major advantages of the reactors are 1. Compact reactor and turbines due to high specific enthalpy of supercritical water 2.Simple plant system because of the once-through coolant cycle 3.Use of the experience of LWR and fossil-fired power plants. The temperatures of the major components such as reactor pressure vessel, coolant pipes, pumps and turbines are within the experience, in spite of the high outlet coolant temperature. 4.Similarity to LWR safety design and criteria, but no burnout phenomenon 5.Potential cost reduction due to smaller material expenditure and short construction period 6.The smallest reactor not in power rating, but in plant sizes. 7.High-thermal efficiency and low coolant flow rate because of high enthalpy rise. 8.Water cooled reactors potentially free from SCC (stress corrosion cracking) problems. 9.Compatibility of tight-fuel-lattice fast reactor core due to small coolant flow rate, potentially easy shift to fast breeder reactor without changing coolant technology. 10.Potential of producing energy products such as hydrogen and high quality hydro carbons. (authors)

  9. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The existing sodium cooled fast reactors (SFR) have two types of designs--loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphenix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL's Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed

  10. Sodium experiment on fully natural circulation systems for decay heat removal in Japan sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The DHRS of JSFR consists of one unit of DRACS (direct reactor auxiliary cooling system), which has a dipped heat exchanger in the reactor vessel and two units of PRACS, which has a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and start-up transient of the DHRS loop with parameters of pressure loss coefficients in the loops. The transient experiments for the start-up of DHRS loop showed that quick increase of natural draft in the air duct followed by smooth increase of sodium flow rate in the DHRS loop. Influences of the pressure loss coefficients in the primary loop and the DHRS loop were limited on the core temperature and also heat removal of PRACS, respectively due to recovery of natural circulation head via the increase of temperature difference in each loop. (author)

  11. Experimental evaluation of a direct air-cooled lithium bromide-water absorption prototype for solar air conditioning

    OpenAIRE

    González-Gil, A.; Izquierdo, M.; Marcos, J.D.; Palacios, E.

    2011-01-01

    Abstract A new direct air-cooled single-effect LiBr-H2O absorption prototype is described and proposed for use in solar cooling. As distinguishing aspects, it presents: an adiabatic absorber using flat-fan sheets; an air-cooling system that directly refrigerates both the condenser and the absorber and; the possibility of being operated also as a double-effect unit. A solar facility comprising a 48m2 field of flat-plate collectors was used to test the single-effect operation mode of...

  12. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Science.gov (United States)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  13. Influence of Different Vortex Generators on Heat Transfer in Direct Air-Cooled Condensers

    Institute of Scientific and Technical Information of China (English)

    ZHOU Guobing; YANG Laishun

    2012-01-01

    With the capacity of air cooling turbines increasing, the air-cooled steam condenser has been developed to the single fiat tube. Fig.1 shows the studied single-row wavy-finned flat tube. Vortex generators are more and more used in oil and chemical industry, power generation, and refrigeration industry in the heat exchanger. The analyses are performed to investigate the effect of vortex generators on pressure drop and heat transfer of single-row wavy-finned fiat tubes of direct air-cooled condensers.

  14. Discussion on Feasibility and Economy of 1000-MW Ultra-Supercritical Air-Cooling Unit

    Institute of Scientific and Technical Information of China (English)

    Zhu Jun; Wang Yunze; Jin Wen

    2007-01-01

    @@ Based on analysis on types and features of domestically made 1000-MW large ultra supercritical steam turbine and 600-MW air cooling steam turbine,the author puts forward that 1000-MW ultra supercritical air-cooling turbine can be assembled with high and medium pressure cylinder modules of 1000-MW ultra-supercritical steam turbine and low-pressure cylinder module of 600-MW tow-cylinder and tow-exhaust air-cooling turbine.In addition,the economy of the assembled turbine is discussed, and designing considerations and issues need to be furtherstudied are proposed as well.

  15. Calculation capability of NETFLOW++ code for natural circulation in sodium cooled fast reactor

    International Nuclear Information System (INIS)

    head produced by cooling, pressure losses and so forth were included. An intentional scram test at 140 MWt at 'Joyo' was conducted to check the function of the heat transport system after the modification in order to increase neutron flux. This result was summarized by Kawahara in an unclassified report and calculated by Takamatsu et al. using the Mimir- N2 code for exclusive use. When the reactor was tripped intentionally, the primary motor was tripped and driven by a pony motor instead. The flow rate in the primary HTS leveled off at approximately 15% of the rated flow rate. Meanwhile, the secondary HTS was circulated by the natural circulation after the secondary pump trip. This event is simulated using a calculation model. The reactor core is divided into 10 subassembly groups from the center to the peripheral. All major components in two loops are taken into account. At time zero, the reactor power is tripped in the calculation and decreased along the decay heat characteristic. At the same time, the pumps in the primary and secondary HTSs in the computer code are tripped and run down based on the coast-down characteristics given by inertias. After a certain period, the primary pumps are operated with the low speed in the same manner as the test using a pump model which can trace the given flow rate automatically. While, the pumps in the secondary HTS are not operated after the trip. Comparison between the measured result and calculation result are shown. The DHX in the figure stands for a dump heat exchanger. The word DHX is used in 'Joyo' instead of the air cooler. Trends with closed symbols and solid lines stand for test result. The open symbols stand for the calculated results. In this case, the same shaped symbol means the same parameter as in the test result. The agreement is the same order as the Mimir-N2 exclusive code for 'Joyo'. As a result, it raise expectations that the NETFLOW++ code is firmly applicable to the natural circulation analysis of sodium-cooled

  16. Experimental results of a direct air-cooled ammonia–lithium nitrate absorption refrigeration system

    International Nuclear Information System (INIS)

    Absorption thermal cooling systems driven by renewable energy are a viable option in order to reduce fossil fuel consumption and the associated emissions. This work shows the results of an air cooled absorption cooling prototype working with an ammonia–lithium nitrate mixture at high ambient temperatures. An absorption refrigeration system was designed and built. The prototype is a one stage ammonia–lithium nitrate air cooled chiller. The experimental system was instrumented to evaluate each component. This paper shows the operation conditions in the experimental unit as well as some of the heat loads encountered at different operating conditions. The system was operated successfully at ambient temperatures in the range of 25–35 °C. A series of test showed that even at ambient temperatures it can be operated at evaporator temperatures below 10 °C producing chilled water for air conditioning applications such as radiative cooling panels. The system proved to stabilize very quickly and no risk of crystallization was encountered so the first results are promising in order to continue with the development of a more advanced prototype. - Highlights: •Experimental results of a direct air-cooled ammonia–lithium nitrate system. •The prototype is a one stage ammonia–lithium nitrate air cooled chiller. •The absorption system was operated successfully at ambient temperatures. •Cooling loads of 4.5 kW were reached in the chilled water side

  17. Investigation on mass flux distribution and asymmetrical cooling in a plate-type fuel reactor

    International Nuclear Information System (INIS)

    A program was developed to calculate the mass flux distribution in the whole core and the asymmetrical cooling of fuel pins for a plate type fuel reactor by applying the proper physical model. Three iterative algorithms for mass flux distribution calculation and two iterative algorithms for temperature field calculation of plate fuel element under asymmetrical cooling condition were proposed and compared by applying in a subassembly. The results showed that the flow distribution is mainly determined by the core structure, although it is also impacted by the power distribution in the core. The asymmetrical cooling condition seriously impacts the temperature field and power distribution of the fuel pins. (authors)

  18. The Fort St. Vrain high temperature gas-cooled reactor. Pt. 9

    International Nuclear Information System (INIS)

    Precited and measured radial region peaking factors differed more than anticipated, most especially in core regions that were susceptible to ingress of diverted cool primary coolant helium into the thermocouple tubes where the individual region outlet temperatures were measured. To compensate for such cooling, inferred region outlet temperatures were empirically determined by parameter comparison to similar regions in the core not subject to ingress of diverted, cool primary coolant. Decalibration (loss of linearity in indicated versus actual reactor power) existed for both the source range (startup) neutron detectors and the power range neutron detectors. In both cases operational restraints were established to compensate for the lack of linearity. (orig.)

  19. Use of Air2Air Technology to Recover Fresh-Water from the Normal Evaporative Cooling Loss at Coal-Based Thermoelectric Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Ken Mortensen

    2009-06-30

    This program was undertaken to build and operate the first Air2Air{trademark} Water Conservation Cooling Tower at a power plant, giving a validated basis and capability for water conservation by this method. Air2Air{trademark} water conservation technology recovers a portion of the traditional cooling tower evaporate. The Condensing Module provides an air-to-air heat exchanger above the wet fill media, extracting the heat from the hot saturated moist air leaving in the cooling tower and condensing water. The rate of evaporate water recovery is typically 10%-25% annually, depending on the cooling tower location (climate).

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

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