WorldWideScience

Sample records for circulation core cooling

  1. Natural Circulation Phenomena on the Cooling Channel of an Ex-vessel Core Catcher

    International Nuclear Information System (INIS)

    The ex-vessel core catcher in an APR1400 is a passive corium cooling system consisting of an inclined engineered cooling channel made of a single channel between the body of the core catcher and the inside wall of the reactor cavity. If the severe accident in a nuclear power plant occurs and the reactor vessel fails, the molten corium ejected from the reactor vessel is relocated in the body of the ex-vessel core catcher. The water from the IRWST is supplied to the engineered cooling channel between the outside of the core catcher body and the reactor cavity wall. The supplied water in the inclined channel should sufficiently remove the decay heat transferred from the corium by boiling off as steam. A buoyancy-driven natural circulation flow through the cooling channel and down-comers is intended to provide effective long-term cooling, and to stabilize thermally the molten corium mixture in the core catcher body. In general, an increase in the natural circulation mass flow rate of the coolant leads to an increase in the critical heat flux (CHF) on the hot wall, thus enhancing the thermal margin. Therefore, it should be ensured and quantified that the water coolant is circulated at a sufficiently high rate through the inclined cooling channel for decay heat removal to maintain the integrity of the ex-vessel core catcher system. The cooling capability of a newly engineered corium cooling system, that is, an ex-vessel core catcher system, was quantified experimentally. The scaling analysis was applied to design the test facility compared with the prototypic core catcher cooling system. The natural circulation flow rates were measured experimentally, and the flow characteristics in the inclined channel were investigated. As the coolant temperature and heat flux increase, the circulation mass flow rate also increases. The experimental results show that the cooling capability of the real core catcher system is sufficient at the given thermal load imposed on the real core

  2. 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

  3. Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors

    International Nuclear Information System (INIS)

    Highlights: • A core flow distribution calculation code for natural circulation LFRs was developed. • The comparison study between the channel method and the CFD method was conducted. • The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted. - Abstract: Small modular natural circulation lead or lead-alloy cooled fast reactor (LFR) is a potential candidate for LFR development. It has many attractive advantages such as reduced capital costs and inherent safety. The core flow distribution calculation is an important issue for nuclear reactor design, which will provide important input parameters to thermal-hydraulic analysis and safety analysis. The core flow distribution calculation of a natural circulation LFR is different from that of a forced circulation reactor. In a forced circulation reactor, the core flow distribution can be controlled and adjusted by the pump power and the flow distributor, while in a natural circulation reactor, the core flow distribution is automatically adjusted according to the relationship between the local power and the local resistance feature. In this paper, a non-uniform heated parallel channel flow distribution calculation code was developed and the comparison study between the channel method and the CFD method was carried out to assess the exactness of the developed code. The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted using the developed code. A core flow distribution optimization design scheme for a 10MW natural circulation LFR was proposed according to the optimization analysis results

  4. Core cooling systems

    International Nuclear Information System (INIS)

    The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

  5. Emergency core cooling device

    International Nuclear Information System (INIS)

    The present invention provides an emergency core cooling device without using a reactor core spray device, in which the reactor core of a BWR type reactor is cooled effectively and certainly by flooding of the reactor core. That is, the emergency core cooling device comprises a high pressure core water injection system as an emergency core cooling system (ECCS) for cooling the inside of the reactor core upon loss of coolants accident (LOCA). By means of the high pressure core water injection system, water is injected from a condensate storage vessel or a suppression pool to the inside of the reactor core shroud upon LOCA. Accordingly, the reactor core is cooled effectively by reactor core flooding. In this device, cooling water can be injected to the inside of the reactor core shroud by means of the high pressure core injection system upon LOCA in which the coolants are discharged from the outside of the reactor core shroud. On the other hand, upon LOCA in which the coolants are discharged from the inside of the reactor core shroud, the cooling water can be supplied to the reactor core by means of a cooling system upon reactor isolation which injects water to the outside of the reactor core or a low pressure water injection system. (I.S.)

  6. Development of core hot spot evaluation method for decay heat removal by natural circulation under transient conditions in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Toward the commercialization of fast reactors, a design study of Japan Sodium Cooled Fast Reactor (JSFR) is being performed. In this design study, adoption of fully natural circulation system is being examined as the decay heat removal system from the viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot for decay heat removal by natural circulation under transient conditions that is necessary for confirming feasibility of the fully natural circulation system. The new method consists of three step analyses in order to include the effects of thermal hydraulic phenomena particular to natural circulation decay heat removal, inter-assembly heat transfer and flow redistribution in fuel assemblies and in the core by buoyancy force, and therefore enables more rational hot spot evaluation rather than conventional ones. The method was applied to an analysis of loss-of-extemal-power event and the result was compared with those by a conventional method and a detailed 3D simulation. It was confirmed that the proposed method can estimate the hot spot with a reasonable degree of conservativeness. (author)

  7. Radio Galaxies in Cooling Cores

    CERN Document Server

    Eilek, J A

    2003-01-01

    A currently active radio galaxy sits at the center of almost every strong cooling core. What effect does it have on the cooling core? Could its effect be strong enough to offset the radiative cooling which should be occuring in these cores? In order to answer these questions we need to know how much energy the radio jet carries to the cooling core; but we have no way to measure the jet power directly. We therefore need to understand how the radio source evolves with time, and how it radiates, in order to use the data to determine the jet power. When some simple models are compared to the data, we learn that cluster-center radio galaxies probably are energetically important -- but not necessarily dominant -- in cooling cores.

  8. 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.

  9. Experimental and Analytical Modeling of Natural Circulation and Forced Circulation BWRs : Thermal-Hydraulic, Core-Wide, and Regional Stability Phenomena

    NARCIS (Netherlands)

    Furuya, M.

    2006-01-01

    Currently, 434 nuclear power plants are in operation worldwide. 21% of them are known as Boiling Water Reactors (BWRs). These BWRs have pumps that cool their reactor cores (the forced circulation BWRs). In the design of new BWRs, ways to cool the core by a natural circulation flow, without pumps, al

  10. Feasibility research of CRDM natural circulation cooling

    International Nuclear Information System (INIS)

    Background: In the second generation pressurized water reactor, the Control Rod Drive Mechanism (CRDM) is mainly cooled by the blast blower, which consumes more energy and is of lower security. Based on the layout of the CRDM group in the Daya Bay nuclear power plant, the EMC-B type of CRDM is taken as research object. Purpose: The temperature distribution of the CRDM group was simulated to verify the feasibility of natural circulation cooling of air. Methods: Several Computational Fluid Dynamics (CFD) based programs were employed for this research. Firstly, Pro/ENGINEER was applied to establish the 3D model of the CRDM group; then the geometrical model was meshed with ICEM; finally, the flow field and temperature distribution were solved by using FLUENT. Results: The temperature field of the CRDM can be divided into three regions, and the temperature of the middle CRDM was highest, while the temperature of the region between the middle and outside regions was lowest due to the relative weak convection. The highest coil temperature is 198℃, below the limit value of 200℃. Conclusion: The CRDM and the coils may be cooled effectively by natural convection of air under given conditions. (authors)

  11. Cool Core Clusters from Cosmological Simulations

    CERN Document Server

    Rasia, E; Murante, G; Planelles, S; Beck, A M; Biffi, V; Ragone-Figueroa, C; Granato, G L; Steinborn, L K; Dolag, K

    2015-01-01

    We present results obtained from a set of cosmological hydrodynamic simulations of galaxy clusters, aimed at comparing predictions with observational data on the diversity between cool-core and non-cool-core clusters. Our simulations include the effects of stellar and AGN feedback and are based on an improved version of the Smoothed-Particle-Hydrodynamics code GADGET-3, which ameliorates gas mixing and better captures gas-dynamical instabilities by including a suitable artificial thermal diffusion. In this Letter, we focus our analysis on the entropy profiles, our primary diagnostic to classify the degree of cool-coreness of clusters, and on the iron profiles. In keeping with observations, our simulated clusters display a variety of behaviors in entropy profiles: they range from steadily decreasing profiles at small radii, characteristic of cool-core systems, to nearly flat core isentropic profiles, characteristic of non cool-core systems. Using observational criteria to distinguish between the two classes of...

  12. Beyond the Cool Core: The Formation of Cool Core Galaxy Clusters

    CERN Document Server

    Burns, J O; Gantner, B; Motl, P M; Norman, M L; Burns, Jack O.; Hallman, Eric J.; Gantner, Brennan; Motl, Patrick M.; Norman, Michael L.

    2006-01-01

    Why do some clusters have cool cores while others do not? In this paper, cosmological simulations, including radiative cooling and heating, are used to examine the formation and evolution of cool core (CC) and non-cool core (NCC) clusters. Numerical CC clusters at z=0 accreted mass more slowly over time and grew enhanced cool cores via hierarchical mergers; when late major mergers occurred, the CCs survived the collisions. By contrast, NCC clusters of similar mass experienced major mergers early in their evolution that destroyed embryonic cool cores and produced conditions that prevent CC re-formation. We discuss observational consequences.

  13. Beyond the Cool Core: The Formation of Cool Core Galaxy Clusters

    Science.gov (United States)

    Burns, J. O.; Hallman, E. J.; Gantner, B.; Motl, P. M.; Norman, M. L.

    Why do some clusters have cool cores while others do not? In this paper, cosmological simulations, including radiative cooling and heating, are used to examine the formation and evolution of cool core (CC) and non-cool core (NCC) clusters. Numerical CC clusters at z=0 accreted mass more slowly over time and grew enhanced cool cores via hierarchical mergers; when late major mergers occurred, the CCs survived the collisions. By contrast, NCC clusters of similar mass experienced major mergers early in their evolution that destroyed embryonic cool cores and produced conditions that prevent CC re-formation. We discuss observational consequences.

  14. Control device for emergency core cooling systems

    International Nuclear Information System (INIS)

    Purpose: To prevent erroneous operations due to repeated start and stop of emergency core cooling systems, as well as control the reactor water level to an appropriate position in the reactor of a BWR type nuclear power plants, in case of loss of coolants accident, in particular, stick open troubles of a releaf valve, by appropriately maintaining the reactor water level. Constitution: Water either from a condensate storage tank or from a pressure suppression chamber is sprayed into a reactor by an emergency core cooling system pump by way of a feedwater line. In the emergency core cooling system, signals prepared by the addition of the flow rate measured by a flowmeter mounted to the releaf valve air exhaust pipe and the flow rate in other exhaust pipe measured by other flowmeter and signals obtained by the flowmeter for the pump exit are inputted into a comparator circuit. The signals therefrom are transmitted to the control device for the emergency core cooling system pump to control the flow rate in the emergency core cooling system. If the flow rate in the relief valve is decreased, the flow rate in the emergency core cooling system is also decreased to equalize the flow rates from and into the core. Thus, the core liquid level can be kept constant, whereby the water inventry is maintained and the safety of the cladding tube is maintained even if the water level system is failed to make the level monitor impossible. (Seki, T.)

  15. Beyond The Cores Of Cool Core Galaxy Clusters

    Science.gov (United States)

    Burns, Jack O.; Hallman, E. J.; Motl, P. M.; Norman, M. L.

    2006-06-01

    We will present the results of cosmological hydro/N-body adaptive mesh refinement simulations in a concordance LCDM cosmology with a peak resolution of approximately 16 kpc. These simulations include radiative cooling, star formation, and supernova feedback. We find that there are very significant differences between cool core (CC) and non-cool core (NCC) galaxy clusters in their properties beyond the cores (r>100 kpc). For example, the shapes and outer slopes of the synthetic X-ray surface brightness and the temperature profiles are strikingly different between NCC and CC clusters. Beta models are poor fits for r>500 kpc in CC clusters leading to inaccurate global mass estimates and strong deviations from scaling relations in contrast to NCC clusters. We will discuss possible explanations involving differences in the local environments in which these clusters form and evolve.

  16. 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

  17. Limitations of detecting inadequate core cooling with core exit thermocouples

    International Nuclear Information System (INIS)

    USNRC has suggested that the thermocouples (TCs) currently installed at the flow exit of a PWR core could be used to detect a condition of inadequate core cooling (ICC). The use of these TCs has been assumed in the USNRC Regulatory Guide 1.97. PWR vendors have responded to this guideline by proposing ICC instrumentation and procedure packages that include the use of core-exit TCs as a principal means of ICC detection. The core-exit TCs are judged to be able to detect an ICC condition because steam in the core will be superheated by the fuel rods and then flow past the TCs during an accident. The detection of superheat in the fluid stream constitutes the indirect detection of a core uncovery and heatup, or ICC. Data have been analyzed from four experiments conducted in the Loss-of-Fluid Test (LOFT) facility and the results indicate that there are two limitations to the detection of ICC by core exit TCs that should be resolved before reliance can be placed in the measurement. The LOFT TCs are described and these limitations are discussed in this paper

  18. Liquid hydrogen target cooled by circulating helium

    International Nuclear Information System (INIS)

    Structure and characteristics of a liquid hydrogen target, where hydrogen is liquefied with liquid helium flow using evaporation heat of liquid helium and vapour cold, are described. Good thermal insulation of liquid helium supply line permits to remove out of the target the most volumetric and heavy component - helium tank - and to supply liquid helium along spreaded pipeline from the Dewar helium flask. It results in considerable reduction of dimensions and weight, the structure simplification and work facilitation with the target. The target having a working volume of 400 mm length and 60 mm diameter was tested. Vacuum casing of the working volume was made of foam plastic, heat flow to the working volume is equal to 1.5 W. Achieving mode of operation including structure cooling and hydrogen liquefaction took approximately 3 h, liquid helium flow rate for liquefaction of 1 l hydrogen is 2.7 l. Liquid helium flow rate in the mode of operation was equal to 0.7 l/h, i.e. target operation period without adding liquid helium to the Dewar flask is 4-5 days. The target described is notable for simplicity in fabrication, reliability in operation and is very suitable for using in experiment as compared to existing targets with hydrogen liquefaction with liquid helium. Unit structure of the target enables to easily change its configuration relative to problems of concrete physical experiment

  19. KAERI Activities on the Cooling Performance of Ex-vessel Core Catcher

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon; Park, Rae Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Wi, Kyung Jin [Chungnam National University, Daejeon (Korea, Republic of); Thanh, Thuy Nguyen Thi [University of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    To improve the thermal margin for the severe accident measures in high-power reactors, engineered corium cooling systems involving boiling-induced two-phase natural circulation have been proposed for decay heat removal. A boiling-induced natural circulation flow is generated in a coolant path between a hot vessel wall and cold coolant reservoir. In general, an increase in the natural circulation mass flow rate of the coolant leads to an increase in the critical heat flux (CHF) on the hot wall, thus enhancing the thermal margin. Recently, a newly engineered corium cooling system, that is, an ex-vessel core catcher, has been considered as one of severe accident mitigation measures for an APR1400. The ex-vessel core catcher in an APR1400 is a passive corium cooling system consisting of an inclined engineered cooling channel made of a single channel between the body of the core catcher and the inside wall of the reactor cavity. If a severe accident in a nuclear power plant occurs and the reactor vessel fails, the molten corium ejected from the reactor vessel is relocated in the body of the ex-vessel core catcher. The water from the IRWST is supplied to the engineered cooling channel between the outside of the core catcher body and the reactor cavity wall. The supplied water in the inclined channel should sufficiently remove the decay heat transferred from the corium by boiling off as steam. A buoyancy-driven natural circulation flow through the cooling channel and down-comers is intended to provide effective long-term cooling, and to thermally stabilize the molten corium mixture in the core catcher body.. In general, an increase in the natural circulation mass flow rate of the coolant leads to an increase in the critical heat flux (CHF) on the hot wall, thus enhancing the thermal margin. Therefore, it should be ensured and quantified that the water coolant is circulated at a sufficiently high rate through the inclined cooling channel for decay heat removal to maintain

  20. Reliability of BWR high pressure core cooling

    International Nuclear Information System (INIS)

    The high pressure coolant injection system (HPCI), and the reactor core isolation cooling system (RCIC) are steam turbine driven systems that can inject water into a boiling water reactor at full operating pressure. Their purpose is to supply water during any failure that allows water to be lost while the reactor is at pressure and temperature. A large number of BWR plants are not meeting HPCI and RCIC performance goals for core cooling. NSAC considers concurrent failure of NPCI and RCIC to be the most probable potential cause of low reactor water level and possibly fuel damage in a boiling water reactor. Between January 1978 and May 1981, 169 licensee event reports were filed where HPCI or RCIC was inoperable or was declared inoperable. The present effort has shown that at least 40% of NPCI and RCIC problems might be averted by a high quality preventive maintenance program. About half of the plants do not perform cold quick-start surveillance testing of HPCI and RCIC. They do perform routine startup tests, but the equipment is first preheated and the startup is relatively gentle. However, emergency start-ups are abrupt and from the cold condition. Therefore, cold quick-start testing is the only way to assure that all components, control systems, and instruments are functioning correctly for automatic safety initiation. (author)

  1. Convective cores in galactic cooling flows

    CERN Document Server

    Kritsuk, A G; Müller, E

    2000-01-01

    We use hydrodynamic simulations with adaptive grid refinement to study the dependence of hot gas flows in X-ray luminous giant elliptical galaxies on the efficiency of heat supply to the gas. We consider a number of potential heating mechanisms including Type Ia supernovae and sporadic nuclear activity of a central supermassive black hole. As a starting point for this research we use an equilibrium hydrostatic recycling model (Kritsuk 1996). We show that a compact cooling inflow develops, if the heating is slightly insufficient to counterbalance radiative cooling of the hot gas in the central few kiloparsecs. An excessive heating in the centre, instead, drives a convectively unstable outflow. We model the onset of the instability and a quasi-steady convective regime in the core of the galaxy in two-dimensions assuming axial symmetry. Provided the power of net energy supply in the core is not too high, the convection remains subsonic. The convective pattern is dominated by buoyancy driven large-scale mushroom-...

  2. Natural circulation in water cooled nuclear power plants: Phenomena, models, and methodology for system reliability assessments

    International Nuclear Information System (INIS)

    In recent years it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. Further, the IAEA Conference on The Safety of Nuclear Power: Strategy for the Future which was convened in 1991 noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to assure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are an ongoing activity in several IAEA Member States. Some new designs also utilize natural circulation as a means to remove core power during normal operation. In response to the motivating factors discussed above, and to foster international collaboration on the enabling technology of passive systems that utilize natural circulation, an IAEA Coordinated Research Project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation was started in early 2004. Building on the shared expertise within the CRP, this publication presents extensive information on natural circulation phenomena, models, predictive tools and experiments that currently support design and analyses of natural circulation systems and highlights areas where additional research is needed. Therefore, this publication serves both to provide a description of the present state of knowledge on natural circulation in water cooled nuclear power plants and to guide the planning and conduct of the CRP in

  3. 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

  4. Why Do Only Some Galaxy Clusters Have Cool Cores?

    OpenAIRE

    Burns, Jack O.; Hallman, Eric J.; Gantner, Brennan; Motl, Patrick M; Michael L. Norman

    2007-01-01

    Flux-limited X-ray samples indicate that about half of rich galaxy clusters have cool cores. Why do only some clusters have cool cores while others do not? In this paper, cosmological N-body + Eulerian hydrodynamic simulations, including radiative cooling and heating, are used to address this question as we examine the formation and evolution of cool core (CC) and non-cool core (NCC) clusters. These adaptive mesh refinement simulations produce both CC and NCC clusters in the same volume. They...

  5. Study of the circulation theory of the cooling system in vertical evaporative cooling generator

    Institute of Scientific and Technical Information of China (English)

    YU; Shunzhou; CAI; Jing; GUO; Chaohong

    2006-01-01

    The article briefly states the current development of evaporative cooling generator and its advantages comparing with generators of traditional cooling. Vertical evaporative cooling generator, which adopts Close-Loop-Self-Cycle with no-pump and free convection boil in the hollow stator bar, is one of the great developments in generator design. This article emphasizes the importance of cooling system in generator; expatiates the circulation theory in two aspects, energy and flow; and analyzes the essential reason,motivity and stability of Close-Loop-Self-Cycle. The article points out that the motivity of the circulation is the heat absorbed by coolant. After absorbing heat the coolant will have the ability of doing work because of the phase change. In another words, it is the buoyancy causing by density difference leads to the Close-Loop-Self-Cycle. This conclusion is validated by experimental data.

  6. CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

    International Nuclear Information System (INIS)

    At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

  7. Helium circulator design considerations for modular high temperature gas-cooled reactor plant

    International Nuclear Information System (INIS)

    Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the US

  8. A GM cryocooler with cold helium circulation for remote cooling

    Science.gov (United States)

    Wang, Chao; Brown, Ethan

    2014-01-01

    A GM cryocooler with new cold helium circulation system has been developed at Cryomech. A set of check valves connects to the cold heat exchanger to convert a small portion of AC oscillating flow in the cold head to a DC gas flow for circulating cold helium in the remote loop. A cold finger, which is used for remote cooling, is connected to the check valves through a pair of 5 m long vacuum insulated flexible lines. The GM cryocooler, Cryomech model AL125 having 120 W at 80 K, is employed in the testing. The cold finger can provide 50 W at 81 K for the power input of 4.1 kW and 70.5 W at 81.8 K for the power input of 6 kW. This simple and low cost design is very attractive for some applications in the near future.

  9. Core debris cooling with flooded vessel or core-catcher. Heat exchange coefficients under natural convection

    International Nuclear Information System (INIS)

    External cooling by natural water circulation is necessary for molten core retention in LWR lower head or in a core-catcher. Considering the expected heat flux levels (between 0.2 to 1.5 MW/m2) film boiling should be avoided. This rises the question of the knowledge of the level of the critical heat flux for the considered geometries and flow paths. The document proposes a state of the art of the research in this field. Mainly small scale experiments have been performed in a very recent past. These experiments are not sufficient to extrapolate to large scale reactor structures. Limited large scale experimental results exist. These results together with some theoretical investigations show that external cooling by natural water circulation may be considered as a reasonable objective of severe accident R and D. Recently (in fact since the beginning of 1994) new results are available from large scale experiments (CYBL, ULPU 2000, SULTAN). These results indicate that CHF larger than 1 MW/m2 can be obtained under natural water circulation conditions. In this report, emphasis is given to the pursuit of finding predictive models for the critical heat flux in large, naturally convective channels with thick walls. This theoretical understanding is important for the capability to extrapolate to different situations (various geometries, flow paths....). The outcome of this research should be the ability to calculate Boundary Layer Boiling situations (2D), channelling boiling situations (1D) and related CHF conditions. However, a more straightforward approach can be used for the analysis of specific designs. Today there are already some CHF data available for hemispherical geometry and these data can be used before a mechanistic understanding is achieved

  10. Modelling of thermohydraulic emergency core cooling phenomena

    International Nuclear Information System (INIS)

    The codes used in the early seventies for safety analysis and licensing were based either on the homogeneous model of two-phase flow or on the so-called separate-flow models, which are mixture models accounting, however, for the difference in average velocity between the two phases. In both cases the behavior of the mixture is prescribed a priori as a function of local parameters such as the mass flux and the quality. The modern best-estimate codes used for analyzing LWR LOCA's and transients are often based on a two-fluid or 6-equation formulation of the conservation equations. In this case the conservation equations are written separately for each phase; the mixture is allowed to evolve on its own, governed by the interfacial exchanges of mass, momentum and energy between the phases. It is generally agreed that such relatively sophisticated 6-equation formulations of two-phase flow are necessary for the correct modelling of a number of phenomena and situations arising in LWR accidental situations. They are in particular indispensible for the analysis of stratified or countercurrent flows and of situations in which large departures from thermal and velocity equilibrium exist. This report will be devoted to a discussion of the need for, the capacity and the limitations of the two-phase flow models (with emphasis on the 6-equation formulations) in modelling these two-phase flow and heat transfer phenomena and/or different core cooling situations. 18 figs., 1 tab., 72 refs

  11. Production circulator fabrication and testing for core flow test loop. Final report, Phase III

    International Nuclear Information System (INIS)

    The performance testing of two production helium circulators utilizing gas film lubrication is described. These two centrifugal-type circulators plus an identical circulator prototype will be arranged in series to provide the helium flow requirements for the Core Flow Test Loop which is part of the Gas-Cooled Fast Breeder Reactor Program (GCFR) at the Oak Ridge National Laboratory. This report presents the results of the Phase III performance and supplemental tests, which were carried out by MTI during the period of December 18, 1980 through March 19, 1981. Specific test procedures are outlined and described, as are individual tests for measuring the performance of the circulators. Test data and run descriptions are presented

  12. Thermal hydraulic analysis of advanced Pb-Bi cooled NPP using natural circulation

    Science.gov (United States)

    Novitrian, Su'ud, Zaki; Waris, Abdul

    2012-06-01

    We present thermal hydraulic analysis for a low power advanced nuclear reactor cooled by lead-bismuth eutectic. In this work is to study the thermal hydraulic analysis of a low power SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) reactor with 125 MWth which a design a core with very small volume and fuel column height, resulting in a negative coolant temperature coefficient and very low channel pressure drop. And also at full power the heat can be completely removed by natural circulation in the primary circuit, thus eliminating the needs for pumps.

  13. Assessment of a core meltdown in the gas-cooled fast breeder reactor with an upflow core

    International Nuclear Information System (INIS)

    This paper discusses the chronological sequence of events and supporting analysis of a postulated total loss of all coolant circulation in the gas-cooled fast breeder reactor (GCFR) with an upflow core. Redundant and diverse cooling systems are provided for decay heat removal, including pressurized natural circulation in the core auxiliary cooling system, which reduce the probability of this postulated event below the range of plant design bases. Nevertheless, this postulated accident has been considered so that the potential for consequence mitigation and containment margin could be investigated. Two distinct phases of the sequence are discussed: (1) the core response to a total loss of forced and natural coolant circulation and (2) the capability of the prestressed concrete reactor vessel (PCRV) to retain molten fuel debris. Specific design features of the GCFR which prevent recriticality and fuel vaporization due to fuel slumping are under investigation. Analytical work has been initiated to determine the potential for consequence mitigation in the PCRV and the containment. Several concepts for postaccident fuel containment have been identified and appear technically feasible

  14. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump

    International Nuclear Information System (INIS)

    Highlights: • The gas-lift pump has been adopted to enhance the natural circulation capability. • LENAC code is developed in my study. • The calculation results by LENAC code show good agreement with experiment results. • Gas mass flow rate, bubble diameter, rising pipe length are important parameters. -- Abstract: The gas-lift pump has been adopted to enhance the natural circulation capability in the type of lead–bismuth alloy cooled reactors such as Accelerator Driven System (ADS) and Liquid–metal Fast Reactor (LMFR). The natural circulation ability and the system safety are obviously influenced by the two phase flow characteristics of liquid metal–inert gas. In this study, LENAC (LEad bismuth alloy NAtural Circulation capability) code has been developed to evaluate the natural circulation capability of lead–bismuth cooled ADS with gas-lift pump. The drift flow theory, void fraction prediction model and friction pressure drop prediction model have been incorporated into LENAC code. The calculation results by LENAC code show good agreement with experiment results of CIRCulation Experiment (CIRCE) facility. The effects of the gas mass flow rate, void fraction, gas quality, bubble diameter and the rising pipe height or the potential difference between heat exchanger and reactor core on natural circulation capability of gas-lift pump have been analyzed. The results showed that in bubbly flow pattern, for a fixed value of gas mass flow rate, the natural circulation capability increased with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow pattern, with the gas mass flow rate increasing, the natural circulation capability initially increased and then declined. And the flow parameters influenced the thermal hydraulic characteristics of the reactor core significantly. The present work is helpful for revealing the law of enhancing the natural circulation capability by gas-lift pump, and providing theoretical

  15. Why Do Only Some Galaxy Clusters Have Cool Cores?

    CERN Document Server

    Burns, Jack O; Gantner, Brennan; Motl, Patrick M; Norman, Michael L

    2007-01-01

    Flux-limited X-ray samples indicate that about half of rich galaxy clusters have cool cores. Why do only some clusters have cool cores while others do not? In this paper, cosmological N-body + Eulerian hydrodynamic simulations, including radiative cooling and heating, are used to address this question as we examine the formation and evolution of cool core (CC) and non-cool core (NCC) clusters. These adaptive mesh refinement simulations produce both CC and NCC clusters in the same volume. They have a peak resolution of 15.6 h^{-1} kpc within a (256 h^{-1} Mpc)^3 box. Our simulations suggest that there are important evolutionary differences between CC clusters and their NCC counterparts. Many of the numerical CC clusters accreted mass more slowly over time and grew enhanced cool cores via hierarchical mergers; when late major mergers occurred, the CC's survived the collisions. By contrast, NCC clusters experienced major mergers early in their evolution that destroyed embryonic cool cores and produced conditions...

  16. Operation practice and implications of circulating cooling water system of American nuclear power plants

    International Nuclear Information System (INIS)

    In this paper, the circulating cooling water system of nuclear power plants (NPP) in United States is summarized, and the operation practices of different cooling water systems, such as once-through, natural and mechanical draft cooling tower, cooling pond, and mixed cooling mode, used by several coastal and inland NPPs are given. Also, based on the related experiences, some suggestions for use of cooling water system in China NPPs are presented. (authors)

  17. Heat transfer in a cooling water pool with tube bundles under natural circulation

    International Nuclear Information System (INIS)

    Highlights: • SMART adopts a passive system to enhance its safety. • Heat transfer tests for the straight tube bundle in the cooling water pool are performed. • Heat transfer is affected by cooling water temperature, and radial location of the tube. • Heat transfer for the tube bundle is slightly high due to turbulence effect. - Abstract: SMART was developed for electricity generation and seawater desalination and adopted a passive system to enhance its safety. This system could passively remove decay heat from the reactor core to the emergency cooldown tank (ECT) through the heat exchanger. A natural circulation flow was established as water covered the tube bundle inside the emergency cooldown tank. Heat transfer tests for the upward straight tube bundle in the emergency cooldown tank were performed to find the characteristics of the passive system design under natural circulation conditions. The heat transfer coefficient at the tube bundle was affected by the cooling water temperature, and the radial location of the tube. However, it has nearly a similar value at the bottom region regardless of the tube location. The average heat transfer coefficient for the tube bundle was slightly higher than that for the single tube owing to the turbulence effect among the tube bundles

  18. Formation of Cool Cores in Galaxy Clusters via Hierarchical Mergers

    CERN Document Server

    Motl, P M; Loken, C; Norman, M L; Bryan, G; Motl, Patrick M.; Burns, Jack O.; Loken, Chris; Norman, Michael L.; Bryan, Greg

    2004-01-01

    We present a new scenario for the formation of cool cores in rich galaxy clusters based on results from recent high spatial dynamic range, adaptive mesh Eulerian hydrodynamic simulations of large-scale structure formation. We find that cores of cool gas, material that would be identified as a classical cooling flow based on its X-ray luminosity excess and temperature profile, are built from the accretion of discrete, stable subclusters. Any ``cooling flow'' present is overwhelmed by the velocity field within the cluster - the bulk flow of gas through the cluster typically has speeds up to about 2,000 km s^-1 and significant rotation is frequently present in the cluster core. The inclusion of consistent initial cosmological conditions for the cluster within its surrounding supercluster environment is crucial when simulating the evolution of cool cores in rich galaxy clusters. This new model for the hierarchical assembly of cool gas naturally explains the high frequency of cool cores in rich galaxy clusters des...

  19. 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.

  20. Simulation of two-phase natural circulation flows at a core catcher experiment facility using the CUPID code

    International Nuclear Information System (INIS)

    A core catcher system has been developed to maintain the integrity of light water reactor containment from molten corium during a severe accident. The system adopts a two-phase natural circulation flow for cooling the molten corium. It has a unique feature of a two-phase flow with the downward-facing heating walls in the inclined cooling channels, which requires in-depth studies for the optimum design and safety analysis of nuclear reactors with a core catcher. Recently, many experimental investigations on the core catcher have been performed. KAERI also has carried out a two-phase natural circulation experiment to evaluate the cooling performance of the core catcher. Under various heat load and flow conditions, the two-phase natural circulation flow was measured. In this study, the core catcher experiments at KAERI are simulated using the CUPID code. The CUPID code has been developed for the thermal–hydraulic analysis of nuclear reactor components, such as reactor vessel, steam generator, containment, etc. It adopts a three-dimensional, transient, three-field model for two-phase flow and includes various physical models and correlations of the interfacial mass, momentum and energy transfer. The results of the three-dimensional simulations show that the CUPID code can realistically represent the two-phase natural circulation flow loop at the facility. (author)

  1. Calculation capability of NETFLOW++ code for natural circulation in sodium cooled fast reactor

    International Nuclear Information System (INIS)

    The present paper describes the simulation of the natural circulation in the secondary heat transport system (HTS) after an intentional plant trip of the experimental fast reactor 'Joyo' with the 140 MWt irradiation core using the plant dynamics analysis code NETFLOW++. This code is an integrated network code to calculate the nuclear steam supply system (NSSS) and the balance of the plant (BOP), i.e., turbine/feedwater system. Up to now, the code has been validated using transient data of the experimental sodium facility PLANDTL, experimental fast reactor 'Joyo' and the prototype fast breeder reactor 'Monju'. These validations are steps to evaluate the natural circulation transient of a large-scale fast breeder reactor. Therefore, the former validation results are introduced to show the degree of agreement. In order to consolidate the applicability of the code to the evaluation of the natural circulation, the present test was selected and simulated using the NETFLOW++ code. Major plant parameters are simulated with good agreement such a similar accuracy as the Mimir-N2 exclusive code for 'Joyo'. As a result, it is concluded that the NETFLOW++ is applicable to the natural circulation analysis of sodium-cooled fast reactors with the similar scale of the prototype reactor 'Monju'. (author)

  2. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    International Nuclear Information System (INIS)

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters

  3. Radial molecular abundances and gas cooling in starless cores

    CERN Document Server

    Sipilä, O

    2012-01-01

    Aims: We aim to simulate radial profiles of molecular abundances and the gas temperature in cold and heavily shielded starless cores by combining chemical and radiative transfer models. Methods: A determination of the dust temperature in a modified Bonnor-Ebert sphere is used to calculate initial radial molecular abundance profiles. The abundances of selected cooling molecules corresponding to two different core ages are then extracted to determine the gas temperature at two time steps. The calculation is repeated in an iterative process yielding molecular abundances consistent with the gas temperature. Line emission profiles for selected substances are calculated using simulated abundance profiles. Results: The gas temperature is a function of time; the gas heats up as the core gets older because the cooling molecules are depleted onto grain surfaces. The contributions of the various cooling molecules to the total cooling power change with time. Radial chemical abundance profiles are non-trivial: different s...

  4. Present status of the RCNP circulation ring comments on magnetized electron cooling

    International Nuclear Information System (INIS)

    After the brief introduction of the RCNP cooling synchrotron (MSR - Multipurpose Storage Ring), the characteristics of the circulation ring (CR), which is now under construction, are summarized. Some efforts toward the MSR, the next stage of the CR, are also described. In the MSR the completely new scheme will be used for beam cooling. The possibility of stimmulated magnetized electron cooling (say, advanced electron cooling), which lies between the ordinary electron cooling and the new scheme, is discussed. (author)

  5. Lead-bismuth cooled reactor with a high level of natural circulation (RBEC-M)

    International Nuclear Information System (INIS)

    The RBEC-M is a lead-bismuth cooled fast reactor with a high level of primary coolant natural circulation and a gas lift system in the primary circuit to ensure a supply of inert gas (argon) in the coolant under the core. The name reflects the basic technology of the concept: a fast neutron spectrum, heavy metal lead-bismuth coolant, a high level of natural circulation with a nominal operation of inert gas blowers and safe cooldown of the core after the trip of gas supply blowers. The RBEC-M reactor is a conceptual development based on the preliminary design of the RBEC reactor, hereafter referred to as a 'basic project'. The direct predecessor of the RBEC-M is the design named RBEC, one of the Russian-developed designs of fast reactors with heavy metal coolants. The preliminary design of the RBEC reactor of 900 MW(th) and 340 MW(e) was completed in the 1990s by Russian design and scientific institutions: OKB 'Gidropress', Russian Research Centre (RRC) 'Kurchatov Institute' and IPPE, with the participation of VNIINM and RIAR. The main objective of the development of the RBEC lead-bismuth cooled fast reactor was to provide a reliable solution for nuclear fuel breeding, while using an approach alternative to sodium cooled fast reactors. It was assumed that design development of a nuclear power plant (NPP) with such reactor could be completed in a rather short period, with modest expenditures for additional testing and qualification of separate equipment units

  6. Cooling history of Earth's core with high thermal conductivity

    Science.gov (United States)

    Davies, Christopher J.

    2015-10-01

    Thermal evolution models of Earth's core constrain the power available to the geodynamo process that generates the geomagnetic field, the evolution of the solid inner core and the thermal history of the overlying mantle. Recent upward revision of the thermal conductivity of liquid iron mixtures by a factor of 2-3 has drastically reduced the estimated power available to generate the present-day geomagnetic field. Moreover, this high conductivity increases the amount of heat that is conducted out of the core down the adiabatic gradient, bringing it into line with the highest estimates of present-day core-mantle boundary heat flow. These issues raise problems with the standard scenario of core cooling in which the core has remained completely well-mixed and relatively cool for the past 3.5 Ga. This paper presents cooling histories for Earth's core spanning the last 3.5 Ga to constrain the thermodynamic conditions corresponding to marginal dynamo evolution, i.e. where the ohmic dissipation remains just positive over time. The radial variation of core properties is represented by polynomials, which gives good agreement with radial profiles derived from seismological and mineralogical data and allows the governing energy and entropy equations to be solved analytically. Time-dependent evolution of liquid and solid light element concentrations, the melting curve, and gravitational energy are calculated for an Fe-O-S-Si model of core chemistry. A suite of cooling histories are presented by varying the inner core boundary density jump, thermal conductivity and amount of radiogenic heat production in the core. All models where the core remains superadiabatic predict an inner core age of ≲ 600Myr , about two times younger than estimates based on old (lower) thermal conductivity estimates, and core temperatures that exceed present estimates of the lower mantle solidus prior to the last 0.5-1.5 Ga. Allowing the top of the core to become strongly subadiabatic in recent times

  7. 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

  8. Why Do Only Some Galaxy Clusters Have Cool Cores?

    Science.gov (United States)

    Burns, Jack O.; Hallman, Eric J.; Gantner, Brennan; Motl, Patrick M.; Norman, Michael L.

    2008-03-01

    Flux-limited X-ray samples indicate that about half of rich galaxy clusters have cool cores. Why do only some clusters have cool cores while others do not? In this paper, cosmological N-body + Eulerian hydrodynamic simulations, including radiative cooling and heating, are used to address this question as we examine the formation and evolution of cool core (CC) and noncool core (NCC) clusters. These adaptive mesh refinement simulations produce both CC and NCC clusters in the same volume. They have a peak resolution of 15.6 h-1 kpc within a (256 h-1 Mpc)3 box. Our simulations suggest that there are important evolutionary differences between CC clusters and their NCC counterparts. Many of the numerical CC clusters accreted mass more slowly over time and grew enhanced CCs via hierarchical mergers; when late major mergers occurred, the CCs survived the collisions. By contrast, NCC clusters experienced major mergers early in their evolution that destroyed embryonic CCs and produced conditions that prevented CC reformation. As a result, our simulations predict observationally testable distinctions in the properties of CC and NCC beyond the core regions in clusters. In particular, we find differences between CC versus NCC clusters in the shapes of X-ray surface brightness profiles, between the temperatures and hardness ratios beyond the cores, between the distribution of masses, and between their supercluster environs. It also appears that CC clusters are no closer to hydrostatic equilibrium than NCC clusters, an issue important for precision cosmology measurements.

  9. Research on enhancement of natural circulation capability in lead-bismuth alloy cooled reactor by using gas-life pump

    International Nuclear Information System (INIS)

    The gas-lift pump has been adopted to enhance the natural circulation capability in the conceptual designs of lead-bismuth alloy cooled reactors such as ADS and LMFR. The natural circulation capability and the system safety have been obviously influenced by the two phase flow characteristics of liquid metal-inert gas. The numerical research was performed to evaluate the natural circulation capability of lead-bismuth alloy cooled ADS with gas-lift pump. Based on the drift-flux flow model, void fraction prediction model and frictional pressure drop prediction model were adopted in the numerical simulation. The effects of the gas mass flow rate, the gas quality, the bubble diameter and the height of rising pipe on natural circulation capability of gas-lift pump were analyzed. The results show that in bubbly flow pattern, for a fixed value of gas mass flow rate, the natural circulation capability increases with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow patterns, with the gas mass flow rate and the gas quality increase, the natural circulation capability increases initially and then decreases. As the height of rising pipe increases, the natural circulation flow rate goes up. The flow parameters influence the thermal hydraulic characteristics of the reactor core significantly. Therefore, in practical engineering application, the gas mass flow rate, gas quality, bubble diameter and rising pipe height are very important parameters for the design of gas-lift pump systems. The present work is helpful for optimizing the design of the natural circulation cooling system by gas-lift pump. (authors)

  10. Passive safety systems and natural circulation in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially improve economics of new nuclear power plant designs. The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future, which was convened in 1991, noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems 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. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA

  11. Examination of AP-1000 Passive Core Cooling System During Lost of Flow Accident

    International Nuclear Information System (INIS)

    The AP – 1000 nuclear power plant is aIII+generation nuclear power plant, having two steam generation loops with 3400 MWthermal powerproduced by 41,448 UO2fuel rods assembled in 157 fuel assemblies and 1170 MW electric power. As aIII+generation nuclear power plant it utilizes passive cooling systems for the core and for the containment. These systems depend solely on natural physical mechanisms such as gravity, natural circulation and compressed gases to generate the driving forces for passive cooling during an accident. These passive systems do not utilize active components such as pumps, fans, chillers and generators to satisfy safety requirements. One of these existing passive systems which are placed in the first cooling loop is calledPassive Residual Heat Removal (PRHR), which takes a crucial role during a Loss of Flow Accident (LOFA). AP-1000 power plant has three major cooling loops, while the first loop is related to the main core cooling system, the second loop is related to the electricity generation loop and the third loop exchanges heat with the environment. During nominal working conditions the first cooling loop includes the core and the steam generator. Through the hot legs pipes, the hot water(coming out of the core) enters the steam generator, exchanges heat with the second loop andreturns through the cold legs pipes to the core entry. While LOFA occurs, the electrical feeds of the main cooling pumps stop and the fluid is stillcirculated, for a short time,as a result of the pumpsinertia,without any cooling water loss.Forced flow cannot exist and natural circulation flow(thermo syphon flow), which is driven by the difference in density along the first cooling loop is being built. At this moment, the PRHRHX, which is immersed in the core storage tank, is passively being connected to the hot leg andis improving the heat removal from the core. The US Nuclear Regulatory Commission (US NRC) licensing defines the safety requirements document for the AP

  12. Buoyant Bubbles and the Disturbed Cool Core of Abell 133

    Science.gov (United States)

    Randall, Scott W.; Clarke, T.; Nulsen, P.; Owers, M.; Sarazin, C.; Forman, W.; Jones, C.; Murray, S.

    2010-03-01

    X-ray cavities, often filled with radio-emitting plasma, are routinely observed in the intracluster medium of clusters of galaxies. These cavities, or "bubbles", are evacuated by jets from central AGN and subsequently rise buoyantly, playing a vital role in the "AGN feedback" model now commonly evoked to explain the balance between heating and radiative cooling in cluster cores. As the bubbles rise, they can displace cool central gas, promoting mixing and the redistribution of metals. I will show a few examples of buoyant bubbles, then argue that the peculiar morphology of the Abell 133 is due to buoyant lifting of cool central gas by a radio-filled bubble.

  13. Tracing Star Formation in Cool Core Clusters with GALEX

    CERN Document Server

    Hicks, Amalia; Donahue, Megan

    2009-01-01

    We present recent results from a GALEX investigation of star formation in 16 cooling core clusters of galaxies, selected to span a broad range in both redshift and central cooling time. Initial results demonstrate clear UV excesses in most, but not all, brightest cluster galaxies in our sample. This UV excess is a direct indication of the presence of young massive stars and, therefore, recent star formation. We report on the physical extent of UV emission in these objects as well as their FUV-NUV colors, and compare GALEX inferred star formation rates to central cooling times, H-alpha and IR luminosities for our sample.

  14. RBMK-1500 accident management for loss of long-term core cooling

    International Nuclear Information System (INIS)

    Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

  15. Radial molecular abundances and gas cooling in starless cores

    OpenAIRE

    Sipilä, O.

    2012-01-01

    Aims: We aim to simulate radial profiles of molecular abundances and the gas temperature in cold and heavily shielded starless cores by combining chemical and radiative transfer models. Methods: A determination of the dust temperature in a modified Bonnor-Ebert sphere is used to calculate initial radial molecular abundance profiles. The abundances of selected cooling molecules corresponding to two different core ages are then extracted to determine the gas temperature at two time steps. The c...

  16. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author)

  17. Magnetorotational instability in cool cores of galaxy clusters

    CERN Document Server

    Nipoti, C; Ettori, S; Bianconi, M

    2015-01-01

    Clusters of galaxies are embedded in halos of optically thin, gravitationally stratified, weakly magnetized plasma at the system's virial temperature. Due to radiative cooling and anisotropic heat conduction, such intracluster medium (ICM) is subject to local instabilities, which are combinations of the thermal, magnetothermal and heat-flux-driven buoyancy instabilities. If the ICM rotates significantly, its stability properties are substantially modified and, in particular, also the magnetorotational instability (MRI) can play an important role. We study simple models of rotating cool-core clusters and we demonstrate that the MRI can be the dominant instability over significant portions of the clusters, with possible implications for the dynamics and evolution of the cool cores. Our results give further motivation for measuring the rotation of the ICM with future X-ray missions such as ASTRO-H and ATHENA.

  18. On the Origin of Cool Core Galaxy Clusters: Comparing X-Ray Observations with Numerical Simulations

    OpenAIRE

    Henning, Jason W.; Gantner, Brennan; Burns, Jack O.; Hallman, Eric J.

    2009-01-01

    To better constrain models of cool core galaxy cluster formation, we have used X-ray observations taken from the Chandra and ROSAT archives to examine the properties of cool core and non-cool core clusters, especially beyond the cluster cores. We produced X-ray images, surface brightness profiles, and hardness ratio maps of 30 nearby rich Abell clusters (17 cool cores and 13 non-cool cores). We show that the use of double beta-models with cool core surface brightness profiles and single beta-...

  19. Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, H.; Kimura, N.; Miyakoshi, H. [Japan Nuclear Cycle Development Institute, Reactor Engineering Group, O-arai Engineering Center, Ibaraki (Japan); Nagasawa, K. [Nuclear Energy System Incorporation, O-arai Office, Ibaraki (Japan)

    2001-07-01

    Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

  20. Preliminary investigation on the primary heat exchanger lower head rupture accident of forced circulation LBE-cooled fast reactor

    International Nuclear Information System (INIS)

    Highlights: • A forced circulation LBE-cooled fast reactor was developed in China. • The steady state of this reactor was simulated by using NTC program. • The HXLHR accident of this reactor was simulated by using NTC program. • Some vapors were dragged into the core by LBE during the HXLHR accident. - Abstract: The problem about the interaction between heavy liquid metal and water is one of the grand challenges in the development of lead or Lead–Bismuth Eutectic (LBE) cooled fast reactor. In this paper, the primary heat exchanger lower head rupture (HXLHR) accident of a forced circulation LBE-cooled fast reactor was simulated with a transient analysis code NTC (Neutronics and Thermal–hydraulics Coupled transient analysis program). The simulation results showed that the water in primary heat exchanger was injected into the primary circuit and vaporized immediately. Then the main vessel was pressurized and the maximum pressure was about 27 bar compared with 0.5 bar in normal condition. During the accident, some of the generated vapors were dragged into the core by LBE, which may cause a reactivity insertion accident. If any positive void coefficient exists in the core, a further study on the HXLHR accident should be performed to evaluate the reactivity insertion accident

  1. 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

  2. Performance and stability analysis of gas-injection enhanced natural circulation in heavy-liquid-metal-cooled systems

    Science.gov (United States)

    Yoo, Yeon-Jong

    The purpose of this study is to investigate the performance and stability of the gas-injection enhanced natural circulation in heavy-liquid-metal-cooled systems. The target system is STAR-LM, which is a 400-MWt-class advanced lead-cooled fast reactor under development by Argonne National Laboratory and Oregon State University. The primary loop of STAR-LM relies on natural circulation to eliminate main circulation pumps for enhancement of passive safety. To significantly increase the natural circulation flow rate for the incorporation of potential future power uprates, the injection of noncondensable gas into the coolant above the core is envisioned ("gas lift pump"). Reliance upon gas-injection enhanced natural circulation raises the concern of flow instability due to the relatively high temperature change in the reactor core and the two-phase flow condition in the riser. For this study, the one-dimensional flow field equations were applied to each flow section and the mixture models of two-phase flow, i.e., both the homogeneous and drift-flux equilibrium models were used in the two-phase region of the riser. For the stability analysis, the linear perturbation technique based on the frequency-domain approach was used by employing the Nyquist stability criterion and a numerical root search method. It has been shown that the thermal power of the STAR-LM natural circulation system could be increased from 400 up to 1152 MW with gas injection under the limiting void fraction of 0.30 and limiting coolant velocity of 2.0 m/s from the steady-state performance analysis. As the result of the linear stability analysis, it has turned out that the STAR-LM natural circulation system would be stable even with gas injection. In addition, through the parametric study, it has been found that the thermal inertia effects of solid structures such as fuel rod and heat exchanger tube should be considered in the stability analysis model. The results of this study will be a part of the

  3. 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.

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

    Science.gov (United States)

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

    2016-05-01

    Managing core temperature is critical to patient outcomes in a wide range of clinical scenarios. Previous devices designed to perform temperature management required a trade-off between invasiveness and temperature modulation efficiency. The Esophageal Cooling Device, made by Advanced Cooling Therapy (Chicago, IL), was developed to optimize warming and cooling efficiency through an easy and low risk procedure that leverages heat transfer through convection and conduction. Clinical data from cardiac arrest, fever, and critical burn patients indicate that the Esophageal Cooling Device performs very well both in terms of temperature modulation (cooling rates of approximately 1.3°C/hour, warming of up to 0.5°C/hour) and maintaining temperature stability (variation around goal temperature ± 0.3°C). Physicians have reported that device performance is comparable to the performance of intravascular temperature management techniques and superior to the performance of surface devices, while avoiding the downsides associated with both. PMID:27043177

  5. Preliminary core design of the European lead-cooled system

    International Nuclear Information System (INIS)

    The design of the European Lead-cooled System (ELSY) is closely linked to GEN-IV initiative wherein Lead-cooled Fast Reactor (LFR) is amongst the six selected options. GEN-IV reactor concepts have to fulfil four strategic goals, namely: sustainability, economics, safety and reliability, proliferation resistance and physical protection. Fast spectrum nuclear reactors, such as ELSY, enable a much more reduction of uranium consumption, produce significantly less MA compared to thermal spectrum facilities and, in a long term, can burn MA more efficiently. To reduce the operational cost it is important to reduce a number of the intermediate reactor shutdowns for the core reshuffling. Therefore the core composition should be designed with the sufficient reactivity reserve and small reactivity swing to assure at least one year of operation without fuel reloading. The paper presents the specifications of the ELSY-600 MWe (1500 MWth) core design and reports the results of a first round of the nuclear physics, thermal-mechanical and thermal-hydraulic assessments of such a core consisting of closed hexagonal MOX-fuel assemblies. It is shown that with a start-up core based on the present fuel rod design and a 2$-reactivity excess, one can operate during 3.5 years without core reshuffling, reaching a peak burnup of about 60 GWd/tHM. Beyond that period pellet-cladding mechanical interaction occurs, requiring a further fuel design optimisation. (authors)

  6. 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

  7. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

  8. Static Instability Analysis of the Natural Circulation Flow in a Passive Containment Cooling System

    International Nuclear Information System (INIS)

    When a severe accident occurs in a nuclear power plant, the containment pressure can increase up to 4 bar and, then, it can threaten the containment's integrity. To avoid such over-pressure in the containment, a passive containment cooling systems (PCCS) has been developed instead of existing active cooling systems. The PCCS can cool down the containment by using a natural circulation flow and, thus, flow instabilities may easily occur. It should be confirmed that both static and dynamic flow instabilities do not occur due to the system characteristics. In this study, mathematical models for the single- and two-phase natural circulation flows in a PCCS are developed. Using the flow models, static instability of the natural circulation flow is investigated. In this study, mathematical models for the single- and two-phase natural circulation flows in a PCCS are developed. Using the models, both single- and two-phase natural circulation flows were investigated in terms of static instability. It is shown that, for both cases, there is only one steady-state mass flow rate which satisfies the integrated momentum equation and the pressure drop along the natural circulation loop increases monotonically according to the mass flow rate. Therefore, it can be said that static instability doesn't exist in the PCCS natural circulation loop. Additional research is needed to investigate the dynamic instability of the PCCS natural circulation flow

  9. 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.

  10. Evaluation method for core thermohydraulics during natural circulation in fast reactors. Numerical predictions of inter-wrapper flow

    International Nuclear Information System (INIS)

    Decay heat removal using natural circulation is one of significant functions for a reactor. As the decay heat removal system, a direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this system, cold sodium is provided in an upper plenum of reactor vessel and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such phenomena was developed, which modeled each subassembly as a rectangular duct with gap region and also the upper plenum. This numerical simulation method was verified by a sodium test and also a water test. We applied this method to the natural circulation in a 600 MWe class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. (author)

  11. Cooling, AGN Feedback and Star Formation in Simulated Cool-Core Galaxy Clusters

    CERN Document Server

    Li, Yuan; Ruszkowski, Mateusz; Voit, G Mark; O'Shea, Brian W; Donahue, Megan

    2015-01-01

    Numerical simulations of active galactic nuclei (AGN) feedback in cool-core galaxy clusters have successfully avoided classical cooling flows, but often produce too much cold gas. We perform adaptive mesh simulations that include momentum-driven AGN feedback, self-gravity, star formation and stellar feedback, focusing on the interplay between cooling, AGN heating and star formation in an isolated cool-core cluster. Cold clumps triggered by AGN jets and turbulence form filamentary structures tens of kpc long. This cold gas feeds both star formation and the supermassive black hole (SMBH), triggering an AGN outburst that increases the entropy of the ICM and reduces its cooling rate. Within 1-2 Gyr, star formation completely consumes the cold gas, leading to a brief shutoff of the AGN. The ICM quickly cools and redevelops multiphase gas, followed by another cycle of star formation/AGN outburst. Within 6.5 Gyr, we observe three such cycles. There is good agreement between our simulated cluster and the observations...

  12. Interplay among Cooling, AGN Feedback and Anisotropic Conduction in the Cool Cores of Galaxy Clusters

    CERN Document Server

    Yang, H -Y K

    2015-01-01

    Feedback from the active galactic nuclei (AGN) is one of the most promising heating mechanisms to circumvent the cooling-flow problem in galaxy clusters. However, the role of thermal conduction remains unclear. Previous studies have shown that anisotropic thermal conduction in cluster cool cores (CC) could drive the heat-flux driven buoyancy instabilities (HBI) that re-orient the field lines in the azimuthal directions and isolate the cores from conductive heating from the outskirts. However, how the AGN interacts with the HBI is still unknown. To understand these interwined processes, we perform the first 3D magnetohydrodynamic (MHD) simulations of isolated CC clusters that include anisotropic conduction, radiative cooling, and AGN feedback. We find that: (1) For realistic magnetic field strengths in clusters, magnetic tension can suppress a significant portion of HBI-unstable modes and thus the HBI is either completely inhibited or significantly impaired, depending on the unknown magnetic field coherence le...

  13. 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

  14. 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

  15. 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.)

  16. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Science.gov (United States)

    Scarlat, Raluca Olga

    This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling

  17. 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

  18. 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

  19. EVERY BCG WITH A STRONG RADIO AGN HAS AN X-RAY COOL CORE: IS THE COOL CORE-NONCOOL CORE DICHOTOMY TOO SIMPLE?

    International Nuclear Information System (INIS)

    The radio active galactic nucleus (AGN) feedback in X-ray cool cores has been proposed as a crucial ingredient in the evolution of baryonic structures. However, it has long been known that strong radio AGNs also exist in 'noncool core' clusters, which brings up the question whether an X-ray cool core is always required for the radio feedback. In this work, we present a systematic analysis of brightest cluster galaxies (BCGs) and strong radio AGNs in 152 groups and clusters from the Chandra archive. All 69 BCGs with radio AGN more luminous than 2 x 1023 W Hz-1 at 1.4 GHz are found to have X-ray cool cores. BCG cool cores can be divided into two classes: the large cool core (LCC) class and the corona class. Small coronae, easily overlooked at z > 0.1, can trigger strong heating episodes in groups and clusters, long before LCCs are formed. Strong radio outbursts triggered by coronae may destroy embryonic LCCs and thus provide another mechanism to prevent the formation of LCCs. However, it is unclear whether coronae are decoupled from the radio feedback cycles as they have to be largely immune to strong radio outbursts. Our sample study also shows the absence of groups with a luminous cool core while hosting a strong radio AGN, which is not observed in clusters. This points to a greater impact of radio heating on low-mass systems than clusters. Few L 1.4GHz > 1024 W Hz-1 radio AGNs (∼16%) host an L 0.5-10keV > 1042 erg s-1 X-ray AGN, while above these thresholds, all X-ray AGNs in BCGs are also radio AGNs. As examples of the corona class, we also present detailed analyses of a BCG corona associated with a strong radio AGN (ESO 137-006 in A3627) and one of the faintest coronae known (NGC 4709 in the Centaurus cluster). Our results suggest that the traditional cool core/noncool core dichotomy is too simple. A better alternative is the cool core distribution function, with the enclosed X-ray luminosity or gas mass.

  20. Operation and Licensing of Mixed Cores in Water Cooled Reactors

    International Nuclear Information System (INIS)

    Nuclear fuel is a highly complex material that is subject to continuous development and is produced by a range of manufacturers. During operation of a nuclear power plant, the nuclear fuel is subject to extreme conditions of temperature, corroding environment and irradiation, and many different designs of fuel have been manufactured with differing fuel materials, cladding materials and assembly structure to ensure these conditions. The core of an operating power plant can contain hundreds of fuel assemblies, and where there is more than a single design of a fuel assembly in the core, whether through a change of fuel vendor, introduction of an improved design or for some other reason, the core is described as a mixed core. The task of ensuring that the different assembly types do not interact in a harmful manner, causing, for example, differing flow resistance resulting in under cooling, is an important part of ensuring nuclear safety. This report has compiled the latest information on the operational experience of mixed cores and the tools and techniques that are used to analyse the core operation and demonstrate that there are no safety related problems with its operation. This publication is a result of a technical meeting in 2011 and a series of consultants meetings

  1. 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)....

  2. Are black holes big enough to quench cooling in cluster cool cores?

    OpenAIRE

    Ruszkowski, M.

    2009-01-01

    Total energy arguments (e.g., Fabian et al. 2002) suggest that black holes need to have masses significantly in excess of the prediction from the classic black hole mass - velocity dispersion relation (M-sigma) in order to offset the cooling losses in massive cool core clusters. This suggests that the black holes may be too small to power such clusters. However, Lauer et al. (2007) argue that the black hole mass - bulge luminosity relationship is a better predictor of black hole masses in hig...

  3. Modeling active galactic nucleus feedback in cool-core clusters: The balance between heating and cooling

    International Nuclear Information System (INIS)

    We study the long-term evolution of an idealized cool-core galaxy cluster under the influence of momentum-driven active galactic nucleus (AGN) feedback using three-dimensional high-resolution (60 pc) adaptive mesh refinement simulations. The feedback is modeled with a pair of precessing jets whose power is calculated based on the accretion rate of the cold gas surrounding the supermassive black hole (SMBH). The intracluster medium first cools into clumps along the propagation direction of the jets. As the jet power increases, gas condensation occurs isotropically, forming spatially extended structures that resemble the observed Hα filaments in Perseus and many other cool-core clusters. Jet heating elevates the gas entropy, halting clump formation. The cold gas that is not accreted onto the SMBH settles into a rotating disk of ∼1011 M ☉. The hot gas cools directly onto the disk while the SMBH accretes from its innermost region, powering the AGN that maintains a thermally balanced state for a few Gyr. The mass cooling rate averaged over 7 Gyr is ∼30 M ☉ yr–1, an order of magnitude lower than the classic cooling flow value. Medium resolution simulations produce similar results, while in low resolution runs, the cluster experiences cycles of gas condensation and AGN outbursts. Owing to its self-regulating mechanism, AGN feedback can successfully balance cooling with a wide range of model parameters. Our model also produces cold structures in early stages that are in good agreement with the observations. However, the long-lived massive cold disk is unrealistic, suggesting that additional physical processes are still needed.

  4. Analysis on Non-Uniform Flow in Steam Generator During Steady State Natural Circulation Cooling

    OpenAIRE

    Susyadi; T. Yonomoto

    2007-01-01

    Investigation on non uniform flow behavior among U-tube in steam generator during natural circulation cooling has been conducted using RELAP5. The investigation is performed by modeling the steam generator into multi channel models, i.e. 9-tubes model. Two situations are implemented, high pressure and low pressure cases. Using partial model, the calculation simulates situation similar to the natural circulation test performed in LSTF. The imposed boundary conditions are flow rate, quality, pr...

  5. Stability of flashing-driven natural circulation in a passive moderator cooling system for Canadian SCWR

    International Nuclear Information System (INIS)

    Highlights: • The stability in a passive moderator cooling system of a unique system in the Canadian SCWR. • Identify and analyze unstable oscillations using flashing-driven natural circulation test results. • The flashing-driven oscillations categorized as a flashing-driven Type-I density wave instability including a geysering-like feature. • A stability map on the dimensionless plane with the Subcooling number and Phase Change number. - Abstract: This paper presents an examination of the instability mechanisms in a Passive Moderator Cooling System for the Canadian SCWR (Supercritical Water-cooled Reactor). The passive system is being developed at AECL using a flashing-driven natural circulation loop. Unstable intermittent and sinusoidal oscillations were identified from experimental data of the flashing-driven natural circulation passive moderator cooling system. The oscillation periods were correlated with the boiling delay time. A stability map for a flashing-driven two-phase natural circulation loop was established on the dimensionless plane with Subcooling number and Phase Change number. It was observed that there is thermal non-equilibrium in the single-phase and two-phase oscillation stages of the flashing-driven natural circulation

  6. Reactivity accidents analysis during natural core cooling operation of ETRR-2

    International Nuclear Information System (INIS)

    One of the main features of Egypt Test and Research Reactor Number 2 (ETRR-2), MTR type, is a continuous steady-state operation at low power level, <=400 kW, with core cooling by natural water circulation. Two flapper valves mounted on the return core cooling pipe lines and long chimney encloses the reactor core and assure natural convection phenomena through the reactor core and reactor pool. Many tests and experiments are carried out during this state of operation. A possible occurrence of reactivity insertion accidents (RIA) may be expected over this operation. The present work studies two types of possible RIA: 1-fast reactivity insertion accident (FRIA) with rate 1.04$/s and 2-slow reactivity insertion accident (SRIA) with rate 0.023$/s which may occur due to fast/slow withdrawal of a control rod or sudden cooling of the core inlet water temperature. Failure or success of the reactor scram system during the transient operation is considered. A computer code TRAP22 is developed for such analysis. It is verified against CONVEC code and commissioning tests for steady state operation. The results of verification show good agreement. The study demonstrates that the reactor can be scrammed safely due to either FRIA or SRIA, whenever the maximum expected hot channel HC clad temperature lies within the range 70.73-71.85 deg. C. While, in case of failure of scram system the maximum (HC) clad temperature reaches the burn out value at time 1.175s for FRIA and at 46.36s for SRIA. At the burn out point the clad surface heat flux exceeds its design critical value which results in partial fuel melt

  7. Evaluation of emergency core cooling system of TAPS-BWR

    International Nuclear Information System (INIS)

    Full text: The twin nuclear reactors (660 MWth) at the Tarapur atomic power station (TAPS) belong to the early generation of boiling water reactors (BWRs). ECC system of TAPS consists of two pairs pipes connected to the main shroud shell. These pipes deliver cold water to core spray ring spargers during emergency cooling following LOCA. In case of loss of coolant accident (LOCA) due to large break in the primary coolant system, depressurization rates are considerably high. This results in the activation of the core spray system, which helps in limiting the plant parameters to safe levels. However, in case of a small break, core spray will take a longer time to come into action due to the large depressurization time. Further in the case of small break, the velocities are also small. This can result in phase separation and hence core heat up. To avoid this an auto relief system is provided in the BWRs at Tarapur, which comes into action by sensing signals from containment pressurization. This system depressurizes the primary coolant system fast. Adequacy of the auto relief system depends on the time at which it comes into action and the discharge rate. This paper deals with the evaluation of the adequacy of the core spray system in conjunction with the auto relief system. The results of analysis of LOCAs caused due to breaks of two different sizes in the recirculation line are presented and discussed

  8. The Impact of Star Formation on Cool Core Galaxy Clusters

    CERN Document Server

    Motl, P M; Norman, M L; Bryan, G L

    2003-01-01

    We present results from recent simulations of the formation and evolution of clusters of galaxies in a LambdaCDM cosmology. These simulations contain our most physically complete input physics to date including radiative cooling, star formation that transforms rapidly cooling material into aggregate star particles and we also model the thermal feedback from resulting supernovae in the star particles. We use an adaptive mesh refinement (AMR) Eulerian hydrodynamics scheme to obtain very high spatial resolution (~ 2 kpc) in a computational volume 256 Mpc on a side with mass resolution for dark matter and star particles of ~ 10^8 M_solar. We examine in detail the appearance and evolution of the core region of our simulated clusters.

  9. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  10. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    International Nuclear Information System (INIS)

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection

  11. Hypercapnia increases core temperature cooling rate during snow burial.

    Science.gov (United States)

    Grissom, Colin K; Radwin, Martin I; Scholand, Mary Beth; Harmston, Chris H; Muetterties, Mark C; Bywater, Tim J

    2004-04-01

    Previous retrospective studies report a core body temperature cooling rate of 3 degrees C/h during avalanche burial. Hypercapnia occurs during avalanche burial secondary to rebreathing expired air, and the effect of hypercapnia on hypothermia during avalanche burial is unknown. The objective of this study was to determine the core temperature cooling rate during snow burial under normocapnic and hypercapnic conditions. We measured rectal core body temperature (T(re)) in 12 subjects buried in compacted snow dressed in a lightweight clothing insulation system during two different study burials. In one burial, subjects breathed with a device (AvaLung 2, Black Diamond Equipment) that resulted in hypercapnia over 30-60 min. In a control burial, subjects were buried under identical conditions with a modified breathing device that maintained normocapnia. Mean snow temperature was -2.5 +/- 2.0 degrees C. Burial time was 49 +/- 14 min in the hypercapnic study and 60 min in the normocapnic study (P = 0.02). Rate of decrease in T(re) was greater with hypercapnia (1.2 degrees C/h by multiple regression analysis, 95% confidence limits of 1.1-1.3 degrees C/h) than with normocapnia (0.7 degrees C/h, 95% confidence limit of 0.6-0.8 degrees C/h). In the hypercapnic study, the fraction of inspired carbon dioxide increased from 1.4 +/- 1.0 to 7.0 +/- 1.4%, minute ventilation increased from 15 +/- 7 to 40 +/- 12 l/min, and oxygen saturation decreased from 97 +/- 1 to 90 +/- 6% (P < 0.01). During the normocapnic study, these parameters remained unchanged. In this study, T(re) cooling rate during snow burial was less than previously reported and was increased by hypercapnia. This may have important implications for prehospital treatment of avalanche burial victims. PMID:14660514

  12. 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.

  13. Design of core cooling monitoring system based on SOP

    International Nuclear Information System (INIS)

    Ling'ao phase Ⅱ nuclear power project is the first SOP adopted plant in China. According to the requirement of this procedure, Core Cooling Monitoring System (CCMS) shall perform two of six status function monitoring tasks of SOP, including primary loop coolant inventory, pressure and temperature. Inventory is monitored by reactor vessel level and the rest are monitored by saturation margin ΔTsat. To fulfill these tasks, the system design, including sensors, data processing and information display, is significantly different from EOP design. This paper gives a generally description from the system design aspect. (authors)

  14. 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.

  15. 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.

  16. 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

  17. Update Knowledge Base for Long-term Core Cooling Reliability

    International Nuclear Information System (INIS)

    This revision of the Knowledge Base for Emergency Core Cooling System Recirculation Reliability (NEA/CSNI/R (95)11) describes the current status (late 2012) of the knowledge base on emergency core cooling system (ECCS) and containment spray system (CSS) suction strainer performance and long-term cooling in operating power reactors. New reactors, such as the AP1000, EPR and APR1400 that are under construction in some Organization for Economic Co-operation and Development (OECD) member countries, are not addressed in detail in this revision. The containment sump (also known as the emergency or recirculation sump in pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs) or the suppression pools or wet wells in boiling water reactors (BWRs)) and associated ECCS strainers are parts of the ECCS in both reactor types. All nuclear power plants (NPPs) are required to have an ECCS that is capable of mitigating a design basis accident (DBA). The containment sump collects reactor coolant, ECCS injection water, and containment spray solutions, if applicable, after a loss-of-coolant accident (LOCA). The sump serves as the water source to support long-term recirculation for residual heat removal, emergency core cooling, and containment atmosphere clean-up. This water source, the related pump suction inlets, and the piping between the source and inlets are important safety-related components. In addition, if fibrous material is deposited at the fuel element spacers, core cooling can be endangered. The performance of ECCS/CSS strainers was recognized many years ago as an important regulatory and safety issue. One of the primary concerns is the potential for debris generated by a jet of high-pressure coolant during a LOCA to clog the strainer and obstruct core cooling. The issue was considered resolved for all reactor types in the mid-1990s and the OECD/NEA/CSNI published report NEA/CSNI/R(95)11 in 1996 to document the state of knowledge of ECCS performance

  18. Investigations on sump cooling after core melt down

    Energy Technology Data Exchange (ETDEWEB)

    Knebel, J.U. [Forschungeszentrum Karlsruhe - Technik und Umwelt Institut fuer Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany)

    1995-09-01

    This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of an optional sump cooling concept for the European Pressurized Water Reactor EPR. This concept is entirely based on passive safety features within the containment. The work is supported by the German utilities and the Siemens dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototypic conditions.

  19. 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

  20. Cool core cycles: Cold gas and AGN jet feedback in cluster cores

    CERN Document Server

    Prasad, Deovrat; Babul, Arif

    2015-01-01

    Using high-resolution 3-D and 2-D (axisymmetric) hydrodynamic simulations in spherical geometry, we study the evolution of cool cluster cores heated by feedback-driven bipolar active galactic nuclei (AGN) jets. Condensation of cold gas, and the consequent enhanced accretion, is required for AGN feedback to balance radiative cooling with reasonable efficiencies, and to match the observed cool core properties. A feedback efficiency (mechanical luminosity $\\approx \\epsilon \\dot{M}_{\\rm acc} c^2$; where $\\dot{M}_{\\rm acc}$ is the mass accretion rate at 1 kpc) as small as $5 \\times 10^{-5}$ is sufficient to reduce the cooling/accretion rate by $\\sim 10$ compared to a pure cooling flow. This value is smaller compared to the ones considered earlier, and is consistent with the jet efficiency and the fact that only a small fraction of gas at 1 kpc is accreted on to the supermassive black hole (SMBH). We find hysteresis cycles in all our simulations with cold mode feedback: {\\em condensation} of cold gas when the ratio...

  1. Method for increasing the stability of a boiling water cooled reactor with natural coolant circulation and a boiling water cooled reactor with natural coolant circulation (its versions)

    International Nuclear Information System (INIS)

    The invention is aimed at improving the safety of a boiling water reactor with natural coolant circulation and increasing the reactor core power density by increasing the coolant flowrate and neutron flux stability as well as by reducing the medium compressibility effectiveness in pressure compensator in dynamic modes. The reactor vessel includes the core, draught section, heat exchangers and a pressure compensator. A part of the pressure compensator is separated by a barrier with calibrated openings possessing a limited capacity and hydrolocks. The calibrated openings in the barrier are located below the coolant level and a part of space separated by a barrier is filled with gas from external system. The part of the barrier projecting above the coolant level is adjacent to heat exchangers. In transitional regimes with the change of pressure in the circulation circuit a hydrolock facilitates to reactor vessel projection against repressing and keeps the barrier from excessive power load

  2. Strong Turbulence in the Cool Cores of Galaxy Clusters: Can Tsunamis Solve the Cooling Flow Problem?

    CERN Document Server

    Fujita, Y; Wada, K

    2004-01-01

    Based on high-resolution two-dimensional hydrodynamic simulations, we show that the bulk gas motions in a cluster of galaxies, which are naturally expected during the process of hierarchical structure formation of the universe, have a serous impact on the core. We found that the bulk gas motions represented by acoustic-gravity waves create local but strong turbulence, which reproduces the complicated X-ray structures recently observed in cluster cores. Moreover, if the wave amplitude is large enough, they can suppress the radiative cooling of the cores. Contrary to the previous studies, the heating is operated by the turbulence, not weak shocks. The turbulence could be detected in near-future space X-ray missions such as ASTRO-E2.

  3. On the Origin of Cool Core Galaxy Clusters: Comparing X-Ray Observations with Numerical Simulations

    CERN Document Server

    Henning, Jason W; Burns, Jack O; Hallman, Eric J

    2009-01-01

    To better constrain models of cool core galaxy cluster formation, we have used X-ray observations taken from the Chandra and ROSAT archives to examine the properties of cool core and non-cool core clusters, especially beyond the cluster cores. We produced X-ray images, surface brightness profiles, and hardness ratio maps of 30 nearby rich Abell clusters (17 cool cores and 13 non-cool cores). We show that the use of double beta-models with cool core surface brightness profiles and single beta-models for non-cool core profiles yield statistically significant differences in the slopes (i.e., beta values) of the outer surface brightness profiles, but similar cluster core radii, for the two types of clusters. Hardness ratio profiles as well as spectroscopically-fit temperatures suggest that non-cool core clusters are warmer than cool core clusters of comparable mass beyond the cluster cores. We compared the properties of these clusters with the results from analogously reduced simulations of 88 numerical clusters ...

  4. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  5. Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    A natural circulation evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500 MW adopting the natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a three-dimensional fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method for the hot spot temperature in the core. The safety analysis method and the three-dimensional analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a sodium test simulating a part of the primary system and the DHRS with about a 1/7 scale, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the three-dimensional analysis adopting the turbulence model. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method. (author)

  6. A Chandra Study of the Image Power Spectra of 41 Cool Core and Non-Cool Core Galaxy Clusters

    CERN Document Server

    Zhang, Chenhao; Zhu, Zhenghao; Li, Weitian; Hu, Dan; Wang, Jingying; Gu, Junhua; Gu, Liyi; Zhang, Zhongli; Liu, Chengze; Zhu, Jie; Wu, Xiang-Ping

    2016-01-01

    In this work we propose a new diagnostic to segregate cool core (CC) clusters from non-cool core (NCC) clusters by studying the two-dimensional power spectra of the X-ray images observed with the Chandra X-ray observatory. Our sample contains 41 members ($z=0.01\\sim 0.54$), which are selected from the Chandra archive when a high photon count, an adequate angular resolution, a relatively complete detector coverage, and coincident CC-NCC classifications derived with three traditional diagnostics are simultaneously guaranteed. We find that in the log-log space the derived image power spectra can be well represented by a constant model component at large wavenumbers, while at small wavenumbers a power excess beyond the constant component appears in all clusters, with a clear tendency that the excess is stronger in CC clusters. By introducing a new CC diagnostic parameter, i.e., the power excess index (PEI), we classify the clusters in our sample and compare the results with those obtained with three traditional C...

  7. Interplay Among Cooling, AGN Feedback, and Anisotropic Conduction in the Cool Cores of Galaxy Clusters

    Science.gov (United States)

    Yang, H.-Y. Karen; Reynolds, Christopher S.

    2016-02-01

    Feedback from the active galactic nuclei (AGNs) is one of the most promising heating mechanisms to circumvent the cooling-flow problem in galaxy clusters. However, the role of thermal conduction remains unclear. Previous studies have shown that anisotropic thermal conduction in cluster cool cores (CCs) could drive the heat-flux-driven buoyancy instabilities (HBIs) that reorient the field lines in the azimuthal directions and isolate the cores from conductive heating from the outskirts. However, how the AGN interacts with the HBI is still unknown. To understand these interwined processes, we perform the first 3D magnetohydrodynamic simulations of isolated CC clusters that include anisotropic conduction, radiative cooling, and AGN feedback. We find the following: (1) For realistic magnetic field strengths in clusters, magnetic tension can suppress a significant portion of HBI-unstable modes, and thus the HBI is either completely inhibited or significantly impaired, depending on the unknown magnetic field coherence length. (2) Turbulence driven by AGN jets can effectively randomize magnetic field lines and sustain conductivity at ∼1/3 of the Spitzer value; however, the AGN-driven turbulence is not volume filling. (3) Conductive heating within the cores could contribute to ∼10% of the radiative losses in Perseus-like clusters and up to ∼50% for clusters twice the mass of Perseus. (4) Thermal conduction has various impacts on the AGN activity and intracluster medium properties for the hottest clusters, which may be searched by future observations to constrain the level of conductivity in clusters. The distribution of cold gas and the implications are also discussed.

  8. 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)

  9. Comparing Cool Cores in the Planck SZ Selected Samples of Clusters of Galaxies with Cool Cores in X-ray Selected Cluster Samples

    Science.gov (United States)

    Jones, Christine; Santos, Felipe A.; Forman, William R.; Kraft, Ralph P.; Lovisari, Lorenzo; Arnaud, Monique; Mazzotta, Pasquale; Van Weeren, Reinout J.; Churazov, Eugene; Ferrari, Chiara; Borgani, Stefano; Chandra-Planck Collaboration

    2016-06-01

    The Planck mission provided a representative sample of clusters of galaxies over the entire sky. With completed Chandra observations of 165 Planck ESZ and cosmology sample clusters at zcore and non-cool core clusters in the Planck-selected clusters with the percentages in X-ray selected cluster samples. We find a significantly smaller percentage of cool core clusters in the Planck sample than in X-ray selected cluster samples. We will discuss the primary reasons for this smaller percentage of cool-core clusters in the Planck-selected cluster sample than in X-ray-selected samples.

  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. Design evaluation of emergency core cooling systems using Axiomatic Design

    International Nuclear Information System (INIS)

    In designing nuclear power plants (NPPs), the evaluation of safety is one of the important issues. As a measure for evaluating safety, this paper proposes a methodology to examine the design process of emergency core cooling systems (ECCSs) in NPPs using Axiomatic Design (AD). This is particularly important for identifying vulnerabilities and creating solutions. Korean Advanced Power Reactor 1400 MWe (APR1400) adopted the ECCS, which was improved to meet the stronger safety regulations than that of the current Optimized Power Reactor 1000 MWe (OPR1000). To improve the performance and safety of the ECCS, the various design strategies such as independency or redundancy were implemented, and their effectiveness was confirmed by calculating core damage frequency. We suggest an alternative viewpoint of evaluating the deployment of design strategies in terms of AD methodology. AD suggests two design principles and the visualization tools for organizing design process. The important benefit of AD is that it is capable of providing suitable priorities for deploying design strategies. The reverse engineering driven by AD has been able to show that the design process of the ECCS of APR1400 was improved in comparison to that of OPR1000 from the viewpoint of the coordination of design strategies

  12. Post-implementation review of inadequate core cooling instrumentation

    International Nuclear Information System (INIS)

    Studies of Three Mile Island (TMI) accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. Industry studies by plant owners and reactor vendors supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC and to provide more complete information for operator control of safety injection, flow to minimize the consequences of such an accident. In 1980, the US Nuclear Regulatory Commission (NRC) required further studies by the industry and described ICC instrumentation design requirements that included human factors and environmental considerations. On December 10, 1982, NRC issued to Babcock and Wilcox (BandW) licensees' orders for Modification of License and transmitted to all pressurized water reactor (PWR) licensees Generic Letter 82-28 to inform them of the revised NRC requirements. The instrumentation requirements for detection of ICC include upgraded subcooling margin monitors (SMMs), upgraded core exit thermocouples (CETs), and installation of a reactor coolant inventory tracking system (RCITS)

  13. 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

  14. PARAMETERS OF WATER CIRCULATION NETWORK FOR A DISTRICT HEATING AND COOLING SYSTEM

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    In a district heating and cooling system, i.e. Beijing combined heating cooling and power (CHCP) system studied here, high temperature water generated by two cogeneration plants circulates through a network between the plants and heat substations. At heat substations, supply water of high temperature from the network drives absorption chillers for air-conditioning in summer and meets space heating demands in winter or domestic hot water demands by heat exchangers in the whole year. The parameters, i.e. supply/return water temperature in the network, has a great impact on primary energy consumption (PEC) of the absorption chillers, circulation pumps and domestic hot water (DHW), which is studied in this paper.

  15. A 3.55 keV line from DM →a→γ: predictions for cool-core and non-cool-core clusters

    Energy Technology Data Exchange (ETDEWEB)

    Conlon, Joseph P.; Powell, Andrew J. [Rudolf Peierls Centre for Theoretical Physics, University of Oxford, 1 Keble Road, Oxford, OX1 3NP (United Kingdom)

    2015-01-13

    We further study a scenario in which a 3.55 keV X-ray line arises from decay of dark matter to an axion-like particle (ALP), that subsequently converts to a photon in astrophysical magnetic fields. We perform numerical simulations of Gaussian random magnetic fields with radial scaling of the magnetic field magnitude with the electron density, for both cool-core 'Perseus' and non-cool-core 'Coma' electron density profiles. Using these, we quantitatively study the resulting signal strength and morphology for cool-core and non-cool-core clusters. Our study includes the effects of fields of view that cover only the central part of the cluster, the effects of offset pointings on the radial decline of signal strength and the effects of dividing clusters into annuli. We find good agreement with current data and make predictions for future analyses and observations.

  16. A 3.55 keV line from $\\text{DM}\\rightarrow a \\rightarrow \\gamma$: predictions for cool-core and non-cool-core clusters

    CERN Document Server

    Conlon, Joseph P

    2014-01-01

    We further study a scenario in which a 3.55 keV X-ray line arises from decay of dark matter to an axion-like particle (ALP), that subsequently converts to a photon in astrophysical magnetic fields. We perform numerical simulations of Gaussian random magnetic fields with radial scaling of the magnetic field magnitude with the electron density, for both cool-core `Perseus' and non-cool-core `Coma' electron density profiles. Using these, we quantitatively study the resulting signal strength and morphology for cool-core and non-cool-core clusters. Our study includes the effects of fields of view that cover only the central part of the cluster, the effects of offset pointings on the radial decline of signal strength and the effects of dividing clusters into annuli. We find good agreement with current data and make predictions for future analyses and observations.

  17. Every BCG with a strong radio AGN has an X-ray cool core

    CERN Document Server

    Sun, Ming

    2009-01-01

    Radio AGN feedback in cool cores has been proposed as a crucial ingredient in the evolution of baryonic structures. However, it has long been known that strong radio AGN also exist in noncool core clusters, which brings up the question whether a cool core is always required for radio feedback. In this work, we present a systematic analysis of BCGs and strong radio AGN in 145 groups and clusters from the Chandra archive. All 65 BCGs with radio AGN more luminous than 2x10^23 W Hz^-1 at 1.4 GHz are found to have X-ray cool cores. The BCG cool cores can be divided into two classes, the large-cool-core (LCC) class and the corona class. Small coronae, easily overlooked at z>0.1, can trigger strong heating episodes in groups and clusters, long before large cool cores are formed. Strong radio outbursts triggered by coronae may destroy embryonic large cool cores and thus provide another mechanism to prevent formation of large cool cores. However, it is unclear whether coronae are decoupled from the radio feedback cycl...

  18. Development of TPNCIRC code for Evaluation of Two-Phase Natural Circulation Flow Performance under External Reactor Vessel Cooling Conditions

    International Nuclear Information System (INIS)

    During a severe accident, corium is relocated to the lower head of the nuclear reactor pressure vessel (RPV). Design concept of retaining the corium inside a nuclear reactor pressure vessel (RPV) through external cooling under hypothetical core melting accidents is called external reactor vessel cooling (ERVC). In this respect, validated two-phase natural circulation flow (TPNC) model is necessary to determine the adequacy of the ERVC design and operating conditions such as inlet area, form losses, gap distance, riser length and coolant conditions. The most important model generally characterizing the TPNC are void fraction and two-phase friction factors. Typical experimental and analytical studies to be referred to on two-phase circulation flow characteristics are those by Reyes, Gartia et al. based on Vijayan et al., Nayak et al. and Dubey et al. In the present paper, two-phase natural circulation (TPNC) flow characteristics under external reactor vessel cooling (ERVC) conditions are studied using two existing TPNC flow models of Reyes and Gartia et al. incorporating more improved void fraction and two-phase friction models. These models and correlations are integrated into a computer program, TPNCIRC, which can handle candidate ERVC design parameters, such as inlet, riser and downcomer flow lengths and areas, gap size between reactor vessel and surrounding insulations, minor loss factors and operating parameters of decay power, pressure and subcooling. Accuracy of the TPNCIRC program is investigated with respect to the flow rate and void fractions for existing measured data from a general experiment and ULPU specifically designed for the AP1000 in-vessel retention. Also, the effect of some important design parameters are examined for the experimental and plant conditions. Using the flow models and correlations are integrated into a computer program, TPNCIRC, a number of correlations have been examined. This seems coming from the differences of void fractions

  19. 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

  20. Post-implementation review of inadequate core cooling instrumentation

    International Nuclear Information System (INIS)

    Studies of the Three Mile Island (TMI) accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. NRC Regulatory Guide 1.97, which covers accident monitoring instrumentation, was revised to be consistent with the requirements of item II.F.2 of NUREG-0737. Following are some of the more significant requirements specified in that item. Instrumentation should provide an unambiguous indication of the approach to and existence of ICC. Reactor water level measurement is to be considered. The system must indicate the existence of ICC caused by various phenomena (e.g., high void fraction pumped flow and stagnant boil-off). The presence of an unrelated phenomenon must not cause the system to erroneously indicate ICC. Advance warning of the approach of ICC must be given. Instrumentation must conform to Appendix B (Class 1E) of NUREG-0737. Alarms and displays should be selected based on a human factors analysis. Instrumentation indications must be integrated into emergency procedures and operator training programs

  1. 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)

  2. General circulation model simulations of recent cooling in the east-central United States

    Science.gov (United States)

    Robinson, Walter A.; Reudy, Reto; Hansen, James E.

    2002-12-01

    In ensembles of retrospective general circulation model (GCM) simulations, surface temperatures in the east-central United States cool between 1951 and 1997. This cooling, which is broadly consistent with observed surface temperatures, is present in GCM experiments driven by observed time varying sea-surface temperatures (SSTs) in the tropical Pacific, whether or not increasing greenhouse gases and other time varying climate forcings are included. Here we focus on ensembles with fixed radiative forcing and with observed varying SST in different regions. In these experiments the trend and variability in east-central U.S. surface temperatures are tied to tropical Pacific SSTs. Warm tropical Pacific SSTs cool U.S. temperatures by diminishing solar heating through an increase in cloud cover. These associations are embedded within a year-round response to warm tropical Pacific SST that features tropospheric warming throughout the tropics and regions of tropospheric cooling in midlatitudes. Precipitable water vapor over the Gulf of Mexico and the Caribbean and the tropospheric thermal gradient across the Gulf Coast of the United States increase when the tropical Pacific is warm. In observations, recent warming in the tropical Pacific is also associated with increased precipitable water over the southeast United States. The observed cooling in the east-central United States, relative to the rest of the globe, is accompanied by increased cloud cover, though year-to-year variations in cloud cover, U.S. surface temperatures, and tropical Pacific SST are less tightly coupled in observations than in the GCM.

  3. Statics and dynamics of a natural circulation cooled boiling water reactor. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Stekelenburg, A.J.C.

    1994-02-21

    Due to the strong interconnection of the various processes in the reactor vessel of a natural circulation cooled boiling water reactor (BWR), explaining the physics of both the statics and the dynamics of the Dodewaard reactor is not an easy task. In this thesis, the physics is studied through a combined experimental and theoretical investigation. The experiments are analyzed further with the use of the model, and the results of the model calculations provide ideas for new experiments. For an experimental study of the reactor behavior, measurement tools are required. Many relevant process variables are supplied by the power plant's data-logger, but a direct method for measuring the circulation flow rate is not available. Reactor behavior can be studied theoreticallly with the use of a complex computer code, based on a multi-node model. In this way, reliable results are obtained. In many cases, however, such a code is not easy to use, and the calculations require much computer time. Calculations based on a simple model have a lower reliability, but, as the model is clearer, provide more insight into the physics of the system. For this reason, a simple theoretical dynamical model for the main physical processes of the Dodewaard natural circulation cooled BWR is presented in the thesis.

  4. On the Formation of Cool, Non-Flowing Cores in Galaxy Clusters via Hierarchical Mergers

    CERN Document Server

    Burns, J O; Norman, M L; Bryan, G L

    2003-01-01

    We present a new model for the creation of cool cores in rich galaxy clusters within a LambdaCDM cosmological framework using the results from high spatial dynamic range, adaptive mesh hydro/N-body simulations. It is proposed that cores of cool gas first form in subclusters and these subclusters merge to create rich clusters with cool, central X-Ray excesses. The rich cool clusters do not possess ``cooling flows'' due to the presence of bulk velocities in the intracluster medium in excess of 1000 km/sec produced by on-going accretion of gas from supercluster filaments. This new model has several attractive features including the presence of substantial core substructure within the cool cores, and it predicts the appearance of cool bullets, cool fronts, and cool filaments all of which have been recently observed with X-Ray satellites. This hierarchical formation model is also consistent with the observation that cool cores in Abell clusters occur preferentially in dense supercluster environments. On the other ...

  5. Availability analysis of the AP600 passive core cooling system

    International Nuclear Information System (INIS)

    The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year and this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). Therefore, it is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3

  6. Why are there strong radio AGNs in the center of "non-cool core" clusters?

    CERN Document Server

    Sun, Ming

    2009-01-01

    Radio AGN feedback in X-ray cool cores has been proposed as a crucial ingredient in the evolution of baryonic structures. However, it has long been known that strong radio AGNs also exist in "noncool core" clusters, which brings up the question whether an X-ray cool core is always required for radio feedback. We present a systematic analysis of 152 groups and clusters to show that every BCG with a strong radio AGN has an X-ray cool core. Those strong radio AGNs in the center of the "noncool core" systems identified before are in fact associated with small X-ray cool cores with typical radii of < 5 kpc (we call them coronae). Small coronae are most likely of ISM origin and they carry enough fuel to power radio AGNs. Our results suggest that the traditional cool core/noncool core dichotomy is too simple. A better alternative is the cool core distribution function with the enclosed X-ray luminosity. Other implications of our results are also discussed, including a warning on the simple extrapolation of the de...

  7. Eastern Pacific cooling and Atlantic overturning circulation during the last deglaciation.

    Science.gov (United States)

    Kienast, Markus; Kienast, Stephanie S; Calvert, Stephen E; Eglinton, Timothy I; Mollenhauer, Gesine; François, Roger; Mix, Alan C

    2006-10-19

    Surface ocean conditions in the equatorial Pacific Ocean could hold the clue to whether millennial-scale global climate change during glacial times was initiated through tropical ocean-atmosphere feedbacks or by changes in the Atlantic thermohaline circulation. North Atlantic cold periods during Heinrich events and millennial-scale cold events (stadials) have been linked with climatic changes in the tropical Atlantic Ocean and South America, as well as the Indian and East Asian monsoon systems, but not with tropical Pacific sea surface temperatures. Here we present a high-resolution record of sea surface temperatures in the eastern tropical Pacific derived from alkenone unsaturation measurements. Our data show a temperature drop of approximately 1 degrees C, synchronous (within dating uncertainties) with the shutdown of the Atlantic meridional overturning circulation during Heinrich event 1, and a smaller temperature drop of approximately 0.5 degrees C synchronous with the smaller reduction in the overturning circulation during the Younger Dryas event. Both cold events coincide with maxima in surface ocean productivity as inferred from 230Th-normalized carbon burial fluxes, suggesting increased upwelling at the time. From the concurrence of equatorial Pacific cooling with the two North Atlantic cold periods during deglaciation, we conclude that these millennial-scale climate changes were probably driven by a reorganization of the oceans' thermohaline circulation, although possibly amplified by tropical ocean-atmosphere interaction as suggested before. PMID:17051216

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

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

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

  9. Physically-Derived Dynamical Cores in Atmospheric General Circulation Models

    Science.gov (United States)

    Rood, Richard B.; Lin, Shian-Kiann

    1999-01-01

    The algorithm chosen to represent the advection in atmospheric models is often used as the primary attribute to classify the model. Meteorological models are generally classified as spectral or grid point, with the term grid point implying discretization using finite differences. These traditional approaches have a number of shortcomings that render them non-physical. That is, they provide approximate solutions to the conservation equations that do not obey the fundamental laws of physics. The most commonly discussed shortcomings are overshoots and undershoots which manifest themselves most overtly in the constituent continuity equation. For this reason many climate models have special algorithms to model water vapor advection. This talk focuses on the development of an atmospheric general circulation model which uses a consistent physically-based advection algorithm in all aspects of the model formulation. The shallow-water model of Lin and Rood (QJRMS, 1997) is generalized to three dimensions and combined with the physics parameterizations of NCAR's Community Climate Model. The scientific motivation for the development is to increase the integrity of the underlying fluid dynamics so that the physics terms can be more effectively isolated, examined, and improved. The expected benefits of the new model are discussed and results from the initial integrations will be presented.

  10. 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)

  11. Results of the reliability analysis of the emergency core cooling system from the Gemeinschaftskernkraftwerk Neckar

    International Nuclear Information System (INIS)

    The appropriate reliability parameter for the assessment of the safety of the emergency core cooling system is the maximum value of the unavailability. For Gemeinschaftskernkraftwerk Neckar power plant, it was possible to demonstrate that by appropriate design it is permissible to transfer normal operating tasks to the emergency cooling system. Moreover, it was possible to give a quantitative basis for the definition of allowable repair times. It is to be noted that all results have been deterministically calculated using the program SAP, developed by Interatom. Description of the emergency core cooling system and its redundancy, presentation of the reliability results (reliability parameters, single cooling chain, total system, repair time)

  12. Aspects of unconventional cores for large sodium cooled power reactors; evaluation of a literature survey

    International Nuclear Information System (INIS)

    The report gives an overview of a literature study on the application of unconventional cores for sodium cooled fast reactors. Different types of unconventional cores (heterogeneous cores, pancake cores, moderated cores and others) are compared with conventional cores, which are characterized by a cylindrical geometry with two or three fissile zones surrounded by an axial and a radial blanket. The main parameters of interest in this comparison are the neutronic parameters sodium void and Doppler effect, the breeding properties and the steel damage. Consequences for the core safety and the overall plant design are also mentioned

  13. The Origin of Ripples in Cool Cores of Galaxy Clusters: Heating by MHD Waves?

    CERN Document Server

    Fujita, Y; Kudoh, T; Yokoyama, T; Fujita, Yutaka; Suzuki, Takeru K.; Kudoh, Takahiro; Yokoyama, Takaaki

    2007-01-01

    We consider MHD waves as a heating source of cool cores of galaxy clusters. In particular, we focus on transverse waves (Alfven waves), because they can propagate a longer distance than longitudinal waves (sound waves). Using MHD simulations, we found that the transverse waves can stably heat a cool core if the wave period is large enough (>~ 10^8 yr). Moreover, the longitudinal waves that are created as a by-product of the nonlinear evolution of the transverse waves could be observed as the 'ripples' found in cool cores.

  14. How unusual is the cool-core radio halo cluster CL1821+643 ?

    CERN Document Server

    Kale, Ruta

    2016-01-01

    Massive galaxy clusters with cool-cores typically host diffuse radio sources called mini-haloes, whereas, those with non-cool-cores host radio haloes. We attempt to understand the unusual nature of the cool-core galaxy cluster CL1821+643 that hosts a Mpc-scale radio halo using new radio observations and morphological analysis of its intra-cluster medium. We present the Giant Metrewave Radio Telescope (GMRT) 610 MHz image of the radio halo. The spectral index, $\\alpha$ defined as $S\\propto \

  15. Are there cool-core clusters at high-redshift? Chandra results and prospects with WFXT

    CERN Document Server

    Santos, Joana S; Rosati, Piero

    2010-01-01

    In this contribution we trace the evolution of cool-core clusters out to z~1.3 using high-resolution Chandra data of three representative cluster samples spanning different redshift ranges. Our analysis is based on the measurement of the surface brightness (SB) concentration, c_SB, which strongly anti-correlates with the central cooling time and allows us to characterize the cool-core strength in low S/N data. We confirm a negative evolution in the fraction of cool-core clusters with redshift, in particular for very strong cool-cores. Still, we find evidence for a large population of well formed cool-cores at z ~ 1. This analysis is potentially very effective in constraining the nature and the evolution of the cool-cores, once large samples of high-z clusters will be available. In this respect, we explore the potential of the proposed mission Wide Field X-ray Telescope (WFXT) to address this science case. We conclude that WFXT provides the best trade-off of angular resolution, sensitivity and covered solid an...

  16. Investigation of corrosion caused by constituents of refinery wastewater effluent used as circulating cooling water.

    Science.gov (United States)

    Zhang, Zhongzhi; Song, Shaofu; Huang, Jie; Ji, Lin; Wu, Fangyun

    2003-01-01

    The corrosion rate of steel plate using single-factor, multifactor, and complex water systems was investigated via refinery wastewater effluents used as circulating cooling water. The results show that the primary corrosion factors of steel depend on the characteristics of the ions, the formation of the oxidized coating, the diffusion of dissolved oxygen, and other complex factors, although ions such as chloride, calcium, and carbonate play an important role. The corrosion rate of carbon steel exhibits two trends: The corrosion rate is high at low conductivity, increases to a maximum, and then decreases and becomes stable with increasing conductivity, as is the case with chloride, sulfate, nitrate and calcium ions. On the other hand, the corrosion rate is highest at low conductivity and then decreases and becomes stable with increasing conductivity, as is the case with carbonate, silicate, and sodium nitrate ions. Research results indicate that the anticorrosive ability is minimal at low conductivity; but is excellent at high conductivity. Pretreatment of low-conductivity water using air flotation and clarification to decrease the concentrations of chloride, calcium, and carbonate ions to a suitable level to satisfy the anticorrosion requirements is required. However, it is not necessary to significantly reduce the salt concentration or conductivity of the water by osmosis or ion exchange to obtain an anticorrosion effect when reusing wastewater effluents as circulating cooling water. PMID:12683464

  17. Thermal-hydraulic analysis for the LBE-cooled natural circulation reactor. Development of the MSG-COPD code and application to the system analysis. Research Document

    International Nuclear Information System (INIS)

    Thermal-hydraulic analysis for the Lead-Bismuth eutectic (LBE)-cooled natural circulation reactor has been conducted by using a combined plant dynamics code (MSG-COPD). MSG-COPD has been developed to consider the multi-dimensional thermal-hydraulics effect on the plant dynamics during transients. Plant dynamics analyses for the LBE-cooled STAR-LM reactor, which has been designed by Argonne National Laboratory in U.S.A., have been performed to understand the basic thermal-hydraulic characteristics of the natural circulation reactor. As a result, it has been made clear that cold coolant remains in the lower plenum by the thermal stratification in case of the ULOHS condition with a severe temperature gradient at the stratified surface in the lower plenum. In addition, the flow-redistribution effect in a core channels by the buoyancy force has been evaluated for a candidate LBE-cooled FBR plant concept (LBE-FR), which has been designed by JNC. A linear evaluation method for the flow-redistribution coefficient is proposed for the LBE-FR, and compared with the multi-dimensional results by MSG-COPD. In conclusion, the method shows sufficient performance for the prediction of the flow-redistribution coefficient for typical lateral power distributions in the core. (author)

  18. Thermal-hydraulic characteristics of a next-generation reactor relying on steam generator secondary side cooling for primary depressurization and long-term passive core cooling

    International Nuclear Information System (INIS)

    System experiments were conducted at the ROSA-V large scale test facility (LSTF) for investigation of new safety systems to mitigate consequences of postulated accidents in pressurized water rectors (PWRs). Tested systems included a steam generator (SG) secondary-side automatic depressurization system (SADS) and gravity-driven injection system (GDIS), which are candidates of safety systems for some next-generation PWR designs. The experimental results showed several thermal-hydraulic behaviors typical of these safety systems, including the primary depressurization due to natural circulation cooling, a nonuniform flow behavior among SG U-tubes, an accumulation of the non-condensable gas originally contained in the injected water, liquid holdup in U-tubes due to the countercurrent flow limiting, and long-term passive core cooling with the GDIS injection. From the assessment of the RELAP5 MOD3 code using the present data, it was found that the inability of the code to predict the U-tube nonuniform flow behavior resulted in overprediction of the primary depressurization rate at a pressure less than 1 MPa, and exaggerated oscillation of the natural circulation flow rate in the primary loop. (orig.)

  19. A Massive, Cooling-Flow-Induced Starburst in the Core of a Highly Luminous Galaxy Cluster

    CERN Document Server

    McDonald, M; Benson, B A; Foley, R J; Ruel, J; Sullivan, P; Veilleux, S; Aird, K A; Ashby, M L N; Bautz, M; Bazin, G; Bleem, L E; Brodwin, M; Carlstrom, J E; Chang, C L; Cho, H M; Clocchiatti, A; Crawford, T M; Crites, A T; de Haan, T; Desai, S; Dobbs, M A; Dudley, J P; Egami, E; Forman, W R; Garmire, G P; George, E M; Gladders, M D; Gonzalez, A H; Halverson, N W; Harrington, N L; High, F W; Holder, G P; Holzapfel, W L; Hoover, S; Hrubes, J D; Jones, C; Joy, M; Keisler, R; Knox, L; Lee, A T; Leitch, E M; Lieu, J; Lueker, M; Luong-Van, D; Mantz, A; Marrone, D P; McMahon, J J; Mehl, J; Meyer, S S; Miller, E D; Mocanu, L; Mohr, J J; Montroy, T E; Murray, S S; Natoli, T; Padin, S; Plagge, T; Pryke, C; Rawle, T D; Reichardt, C L; Rest, A; Rex, M; Ruhl, J E; Saliwanchik, B R; Saro, A; Sayre, J T; Schaffer, K K; Shaw, L; Shirokoff, E; Simcoe, R; Song, J; Spieler, H G; Stalder, B; Staniszewski, Z; Stark, A A; Story, K; Stubbs, C W; Suhada, R; van Engelen, A; Vanderlinde, K; Vieira, J D; Vikhlinin, A; Williamson, R; Zahn, O; Zenteno, A

    2012-01-01

    In the cores of some galaxy clusters the hot intracluster plasma is dense enough that it should cool radiatively in the cluster's lifetime, leading to continuous "cooling flows" of gas sinking towards the cluster center, yet no such cooling flow has been observed. The low observed star formation rates and cool gas masses for these "cool core" clusters suggest that much of the cooling must be offset by astrophysical feedback to prevent the formation of a runaway cooling flow. Here we report X-ray, optical, and infrared observations of the galaxy cluster SPT-CLJ2344-4243 at z = 0.596. These observations reveal an exceptionally luminous (L_2-10 keV = 8.2 x 10^45 erg/s) galaxy cluster which hosts an extremely strong cooling flow (dM/dt = 3820 +/- 530 Msun/yr). Further, the central galaxy in this cluster appears to be experiencing a massive starburst (740 +/- 160 Msun/yr), which suggests that the feedback source responsible for preventing runaway cooling in nearby cool core clusters may not yet be fully establishe...

  20. The Impact of Star Formation on Cool Core Galaxy Clusters

    OpenAIRE

    Motl, P. M.; Burns, J. O.; Norman, M. L.; Bryan, G L

    2003-01-01

    We present results from recent simulations of the formation and evolution of clusters of galaxies in a LambdaCDM cosmology. These simulations contain our most physically complete input physics to date including radiative cooling, star formation that transforms rapidly cooling material into aggregate star particles and we also model the thermal feedback from resulting supernovae in the star particles. We use an adaptive mesh refinement (AMR) Eulerian hydrodynamics scheme to obtain very high sp...

  1. Reliability analysis of passive emergency cooling system of WWER-440 type core

    International Nuclear Information System (INIS)

    The passive emergency core cooling system of WWER-440 reactors consists of two independent identical subsystems, each comprising two similar presurized containers connected via discharge pipes to the reactor mixing chambers, one to the top chamber, the other to the bottom one. The system action starts autonomously upon pressure drop in the primary circuit below 6 MPa when the boric acid solution is forced out to the mixing chambers of the reactor by nitrogen overpressure. The system is activated upon, e.g., main circulating pipe rupture within roughly 10 seconds and its action is necessary as long as around 30 seconds following the accident. An analysis is made of possible system failures and their causes. The failure tree method was used in assessing the system reliability. It was shown that the following events mostly affected its reliability: a fault in container pressure measurement and display, a leak of the container safety valve, and the failure of the back valve in the discharge pipe to open. The analysis also showed that only under the condition when all four containers are operation worthy the reliability is satisfactory. In unit operation with one container out of operation for repair, the system reliability markedly decreased. The existing regulations that have so far permitted the shutdown of one container for three days should be amended to this effect. (Z.M.). 5 figs., 6 refs

  2. 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

  3. Cooling history of Earth's core with high thermal conductivity

    OpenAIRE

    Davies, CJ

    2015-01-01

    Thermal evolution models of Earth's core constrain the power available to the geodynamo process that generates the geomagnetic field, the evolution of the solid inner core and the thermal history of the overlying mantle. Recent upward revision of the thermal conductivity of liquid iron mixtures by a factor of 2-3 has drastically reduced the estimated power available to generate the present-day geomagnetic field. Moreover, this high conductivity increases the amount of heat that is conducted o...

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

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

  5. 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)

  6. Analysis on Non-Uniform Flow in Steam Generator During Steady State Natural Circulation Cooling

    Directory of Open Access Journals (Sweden)

    Susyadi

    2007-07-01

    Full Text Available Investigation on non uniform flow behavior among U-tube in steam generator during natural circulation cooling has been conducted using RELAP5. The investigation is performed by modeling the steam generator into multi channel models, i.e. 9-tubes model. Two situations are implemented, high pressure and low pressure cases. Using partial model, the calculation simulates situation similar to the natural circulation test performed in LSTF. The imposed boundary conditions are flow rate, quality, pressure of the primary side, feed water temperature, steam generator liquid level, and pressure in the secondary side. Calculation result shows that simulation using model with nine tubes is capable to capture important non-uniform phenomena such as reverse flow, fill-and-dump, and stagnant vertical stratification. As a result of appropriate simulation of non uniform flow, the calculated steam generator outlet flow in the primary loop is stable as observed in the experiments. The results also clearly indicate the importance of simulation of non-uniform flow in predicting both the flow stability and heat transfer between the primary and secondary side. In addition, the history of transient plays important role on the selection of the flow distribution among tubes. © 2007 Atom Indonesia. All rights reserved

  7. Experimental Study on Heat Transfer Enhancement of Natural Circulation Liquid Cooling System for Electronic Component

    Institute of Scientific and Technical Information of China (English)

    张正国; 李倩侠; 方晓明; 本田博司

    2004-01-01

    The present research is an experimental study on heat transfer characteristics of a natural circulation cooling system for electronic components. A smooth chip and two micro-pin-finned chips were tested. The chip is mounted on the base of a rectangular horizontal duct located at the bottom of 250 mm high natural circulation loop.FC-72 is used as a coolant. The test conditions are set that the operation pressure of experimental system is 1. 013× 105 Pa, the flow rate of FC-72 is 150 g/min and the subcoolings are 10 K, 25 K and 35 k, respectively. Effect of the subcooling on nucleate boiling and critical heat flux(CHF) were investigated. The results show that subcoolingis found to significantly affect CHF for all chips and micro-pin-finned chips sharply enhanced the boiling heat transfer, CHF of micro-pin-finned chips are 2.5~3 times as large as that of smooth chip at the same subcooling.

  8. Oxygen concentration diffusion analysis of lead-bismuth-cooled, natural-circulation reactor

    International Nuclear Information System (INIS)

    The feasibility study on fast breeder reactors in Japan has been conducted at JNC and related organizations. The Phase-I study has finished in March, 2001. During the Phase-I activity, lead-bismuth eutectic coolant has been selected as one of the possible coolant options and a medium-scale plant, cooled by a lead-bismuth natural circulation flow was studied. On the other side, it is known that lead-bismuth eutectic has a problem of structural material corrosiveness. It was found that oxygen concentration control in the eutectic plays an important role on the corrosion protection. In this report, we have developed a concentration diffusion analysis code (COCOA: COncentration COntrol Analysis code) in order to carry out the oxygen concentration control analysis. This code solves a two-dimensional concentration diffusion equation by the finite differential method. It is possible to simulate reaction of oxygen and hydrogen by the code. We verified the basic performance of the code and carried out oxygen concentration diffusion analysis for the case of an oxygen increase by a refueling process in the natural circulation reactor. In addition, characteristics of the oxygen control system was discussed for a different type of the control system as well. It is concluded that the COCOA code can simulate diffusion of oxygen concentration in the reactor. By the analysis of a natural circulation medium-scale reactor, we make clear that the ON-OFF control and PID control can well control oxygen concentration by choosing an appropriate concentration measurement point. In addition, even when a trouble occurs in the oxygen emission or hydrogen emission system, it observes that control characteristic drops away. It is still possible, however, to control oxygen concentration in such case. (author)

  9. On 135Xe poisoning in the core of a thermal reactor with circulating fuel

    International Nuclear Information System (INIS)

    The derivation of simple analytical expressions for estimating 135Xe poisoning in quasistationary state of the reactor with circulating fuel in the primary circuit. It is shown that 135Xe poisoning in such reactors depends on the ratio of the time during which fuel stays inside the core to the time outside the core (t1/t2).Even at ratio t1/t2=0.1, xenon poisoning effect can the reduced by six times compared to the reactor with fixed fuel, which essentially increases fuel use efficiency

  10. 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.)

  11. Study of risk reduction by improving operation of reactor core isolation cooling system

    International Nuclear Information System (INIS)

    The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

  12. Transmission of waste heat to the environment - cooling with river-water and in circulating systems

    International Nuclear Information System (INIS)

    There is at present in the Federal Republic a revolution in the application of cooling methods, due to the present water economy situation for cooling water supply. Until the end of the 60's fresh-water cooling governed; today, wet closed-circuit cooling in cooling towers is coming through. Furthermore, the application of dry cooling required for the future is being prepared. A survey of the cooling methods, the related problems and the economic effects is given. (orig.)

  13. The thermal structure of the cool core in the Phoenix cluster

    Science.gov (United States)

    Tozzi, Paolo

    2012-10-01

    The SZ-selected cluster SPT-CLJ2344-4243 at z~0.56 (the Phoenix cluster) shows for the first time a hint of a massive cooling-flow-induced starburst, suggesting that the feedback source responsible for preventing runaway cooling may not yet be fully established. We propose to robustly estimate the emission measure distribution of the cool core in the Phoenix cluster, and its temperature and abundance profiles out to 500 kpc, with a medium-deep (210 ks) EPIC observation, in order to investigate the actual structure of the cool core. The proposed study will provide secure science results with a relatively modest exposure, paving the way to an eventual deeper observation of this exceptional and puzzling source.

  14. Penetrating Gas Streams Generate Unrelaxed,Non-Cool-Core Clusters of Galaxies

    CERN Document Server

    Zinger, E; Birnboim, Y; Kravtsov, A; Nagai, D

    2015-01-01

    We utilize cosmological simulations of 16 galaxy clusters at redshifts $z=0$ and $z=0.6$ to study the effect of inflowing streams on the properties of the inner Intra-Cluster Medium (ICM). We find that the mass accretion occurs predominantly along streams that originate from the cosmic web and consist of heated gas. Clusters that are unrelaxed in terms of their X-ray morphology are characterized by higher mass inflow rates and deeper penetration of the streams, typically into the inner third of the virial radius. The penetrating streams generate elevated random motions, bulk flows, cold fronts and metal mixing, thus producing Non-Cool-Core clusters. The degree of penetration of the streams may change over time such that clusters can switch from being unrelaxed to relaxed over a time-scale of several Gyrs. The stream properties thus help us understand the distinction between cool-core and non-cool-core clusters.

  15. FARM: a new tool for optimizing the core performance and safety characteristics of gas cooled fast reactor cores

    International Nuclear Information System (INIS)

    Designing and optimising a reactor core is rather complex as it involves neutronics, thermal-hydraulics and thermomechanics. In order to tentatively overcome these difficulties, a new approach based on simplified models, is being developed aiming in optimising both core performance (core volume, in-cycle Pu inventory..) and core safety characteristics (neutronics coefficients, core pressure drop, transient response..) of a Fast Neutron Reactor. This new approach, called FARM (Fast Reactor Methodology) is currently used for studying a Helium-Cooled Fast Reactor core with carbide fuel pins, and a SiC-based CMC (Ceramic Matrix Composite) cladding. This method has demonstrated that, for a given initial set of specifications (thermal power, inlet coolant temperature, He pressure), 10 optimization variables are sufficient to estimate fair core design features. All simplified models are built from reference CEA codes (ERANOS for neutronics, METEOR for fuel thermomechanics) by way of polynomial interpolations derived from physical analytical considerations. Some safety aspects are also considered in the analysis using analytical descriptions (decay heat removal by natural convection, thermal inertia of the core, etc...). With a multi-criterion genetic algorithm, the 10 optimization variables are then searched for improving both neutronics and safety characteristics. This new methodology allows less accurate, but optimized, core design features to be obtained and proves they are the best that fulfil all the requirements. The first series of studies justify several safety trends already considered in the conventional method (minimisation of pressure drop). Current results confirm that such an approach is possible, and leads to new core designs, similar to the reference core, but with better performance (at least, supply pumping power reduced by 30%, for the same core performance). (authors)

  16. Study of cylindrical nuclear heat source internally cooled by liquid metal circulation

    International Nuclear Information System (INIS)

    In view of fuel management, the early adoption of 14-month fuel shuffling in PWR was a dramatically long refueling period compared to that of the fossil power plant. It is well known that the material integrity of the fuel and the associated structures is the key variable in limiting the fuel residence time in the reactor core. With this view, if one recall that one of advantages of nuclear fuel over the fossil fuel is to convert the fertile material into the fissile while the fuel is burning, it is possible to design reactor core of very long life, that is, more that about 10 years as long as the structural integrity of the fuel is guaranteed. The idea of the long life core may significantly diversify nuclear power source utilization since we may solve many problems in the areas of fuel handling and transportation, waste disposition, reactor operation, nuclear proliferation. In this study, an idea of secure nuclear heat source has been presented with the assumption of possible long-life core. The nuclear heat source of cylindrical shape includes the fuel, the coolant and all the other control devices in a single module which can be manufactured and assembled in a factory. With simple installation effort on the site, the cylindrical nuclear heat source can be augmented with the secondary system which is designed to convert the heat into the other useful form of energy. The design analysis models have been developed for the presented generic structure of the liquid metal cooled reactor of cylinder. The size, power rating and the other important design parameters have been presented. (author)

  17. Modular high-temperature gas-cooled reactor core heatup accident simulations

    International Nuclear Information System (INIS)

    The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs

  18. On the connection between radio mini-halos and gas heating in cool core clusters

    CERN Document Server

    Bravi, Luca; Brunetti, Gianfranco

    2016-01-01

    In this work, we present a study of the central regions of cool-core clusters hosting radio mini-halos, which are di use synchrotron sources extended on cluster-scales surrounding the radio-loud brightest galaxy. We aim to investigate the interplay between the thermal and non-thermal components in the intracluster medium in order to get more insights into these radio sources, whose nature is still unclear. It has recently been proposed that turbulence plays a role for heating the gas in cool cores. A correlation between the radio luminosity of mini-halos, $\

  19. Performance studies of a new core cooling monitor in a boiling water reactor

    International Nuclear Information System (INIS)

    The paper describes the performance studies of a new core cooling monitor (electrical cylindrical heater) for BWRs. Such a detector has been successfully tested at various elevations, including the lower plenum, in the Barsebaeck nuclear power plant under normal operating conditions, and also in various environments in a 160 bar loop (with sudden uncoveries) and in the laboratory (up to 1265 C). It can be operated in two modes: the core cooling mode and the temperature mode, where it actually acts as a thermometer. It currently appears ready for implementation in BWR installations. (orig.)

  20. Fundamental design bases for independent core cooling in Swedish nuclear power reactors

    International Nuclear Information System (INIS)

    New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

  1. The Origin of Ripples in Cool Cores of Galaxy Clusters: Heating by MHD Waves?

    OpenAIRE

    Fujita, Yutaka; Suzuki, Takeru K.; Kudoh, Takahiro; Yokoyama, Takaaki

    2007-01-01

    We consider MHD waves as a heating source of cool cores of galaxy clusters. In particular, we focus on transverse waves (Alfven waves), because they can propagate a longer distance than longitudinal waves (sound waves). Using MHD simulations, we found that the transverse waves can stably heat a cool core if the wave period is large enough (>~ 10^8 yr). Moreover, the longitudinal waves that are created as a by-product of the nonlinear evolution of the transverse waves could be observed as the ...

  2. Integral testing of the AP600 passive emergency core cooling systems

    International Nuclear Information System (INIS)

    Its support of the development of AP600, Westinghouse is conducting two integral systems tests to examine the performance of the passive safety systems. A full-height, full pressure test is being performed to simulate a small loss-of-coolant, steam generator tube rupture and large steam line break events. A one-quarter scale, low pressure test is being performed to simulate transients with emphasis on the transition to the natural circulation post-accident, long-term cooling mode and to demonstrate the long-term cooling capability. Each of the tests will provide detailed experimental results for verification of the accident analysis computer codes. (Author)

  3. Thermal-hydraulic evaluation study of the effectiveness of emergency core cooling system for light water reactors

    International Nuclear Information System (INIS)

    In order to evaluate the core cooling capability of the emergeny core cooling system, which is a safety guard system of light water reactors for a loss-of-coolant accident, a variety of large scale test were performed. Through the results, many phenomena were investigated and the predictabity of analytical codes were examined. The tests conducted were a single-vessel blowdown test, emergency core cooling test in a PWR simulation facility, spray cooling test for a BWR, large scale reflood test and a separate effect test on countercurrent flow. These test results were examined to clarify thermal-hydraulic phenomena and the effect of various test parameters and were utilized to improve predictability of the analytical codes. Some models for flow behavior in the upper core were also developed. By evaluating the effectiveness of various emergency core cooling system configurations, more effective cooling system than the current one was proposed and demonstrated. (author)

  4. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  5. Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

    Directory of Open Access Journals (Sweden)

    M. Vaisnoras

    2008-05-01

    Full Text Available The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.

  6. Alcohol lowers the vasoconstriction threshold in humans without affecting core cooling rate during mild cold exposure.

    Science.gov (United States)

    Johnston, C E; Bristow, G K; Elias, D A; Giesbrecht, G G

    1996-01-01

    Elevated blood alcohol levels are often seen in hypothermia and hyperthermia related deaths, leading to the belief that alcohol renders humans poikilothermic. We examined the core temperature (Tco) thresholds for sweating, vasoconstriction and shivering as well as core cooling rates of seven subjects immersed in 28 degrees C water. On two separate days, subjects exercised on an underwater cycle ergometer to elevate Tco above the sweating threshold. They then rested and cooled until they shivered vigorously. Subjects drank orange juice (7 ml.kg-1) prior to immersion during the control trial and 1 ml.kg-1 absolute ethanol, added to orange juice in a 1:6 ratio, during the alcohol trial. Mean blood alcohol concentration (breath analysis) was 0.097 +/- 0.010 g% at the start of cooling and 0.077 +/- 0.008 g% at the end of the cooling period. Alcohol lowered the vasoconstriction threshold by 0.32 +/- 0.2 degrees C and elevated finger tip blood flow, but had no effect on thresholds for sweating and shivering or core cooling rate. Considering these minor effects it is unlikely that moderate alcohol consumption predisposes individuals to hypothermia or hyperthermia via impaired thermoregulation, but rather likely due to behavioral factors. PMID:8897037

  7. Feasibility study for core cooling performance using SG secondary-side depressurization in PWR

    International Nuclear Information System (INIS)

    In light of the lessons learned from station blackout accidents of the Fukushima Dai-ichi reactor, it is important to line up various cooling measures for reactor core and containment. We are progressing to develop a reliable alternative safety measure to cool the reactor core under small break loss-of-coolant accident (SBLOCA) of PWR using SG secondary-side depressurization. In this research, we aim to promote an early activation of accumulators (ACC) and low-pressure injection (LPI) system to assure the core cooling by an early SG secondary-side depressurization even under loss of core cooling functions by high-pressure injection system. The feasibility study of the safety measure then is being performed by the ROSA / large-scale test facility (LSTF), where tests can be conducted under full-pressure, at Japan Atomic Energy Agency since 2011. The applicability of safety evaluation code M-RELAP5 is also being investigated to establish an evaluation technique for an actual reactor. In this paper, we will present the outline of the safety measure, typical test results and M-RELAP5 calculation results. It is confirmed that the new safety measure is feasible and M-RELAP5 can apply to the SBLOCA transients. (author)

  8. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... on June 7, 2011 (76 FR 32878), for a 60-day public comment period. The public comment period closed... published for public comment on June 15, 2012 (77 FR 36014). A total of 45 comments were received on DG-1277... COMMISSION Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

  9. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH1.8) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  10. Heat pipe based passive emergency core cooling system for safe shutdown of nuclear power reactor

    International Nuclear Information System (INIS)

    On March 11th, 2011, a natural disaster created by earthquakes and Tsunami caused a serious potential of nuclear reactor meltdown in Fukushima due to the failure of Emergency Core Cooling System (ECCS) powered by diesel generators. In this paper, heat pipe based ECCS has been proposed for nuclear power plants. 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 × 10 m × 5 m naturally cooled finned condenser installed outside the primary containment. Heat pipe with overall thermal resistance of 1.44 × 10−5 °C/W will be able to reduce reactor temperature from initial working temperature of 282 °C to below 250 °C within 7 h. The overall ECCS also includes feed water flooding of the core using elevated water tank for initial 10 min which will accelerate cooling of the core, replenish core coolant during loss of coolant accident and avoids heat transfer crisis phenomena during heat pipe start-up process. The proposed heat pipe system will operate in fully passive mode with high runtime reliability and therefore provide safer environment to nuclear power plants. - Highlights: • Completely passive emergency core cooling system (ECCS) for nuclear power plants. • ECCS consists of loop type heat pipe and initial feed water flooding system. • Overall thermal resistance of loop type heat pipe is 1.44 × 10−5 °C/W. • Heat pipe system can reduce reactor temperature from 282 °C to 250 °C in 7 h. • Proposed system will provide reliable and safer cooling for nuclear reactor

  11. Scaling laws and design aspects of a natural-circulation-cooled simulated boiling water reactor fuel assembly

    International Nuclear Information System (INIS)

    In order to study the thermohydraulic behavior of a natural-circulation-cooled boiling water reactor (BWR) fuel assembly, such as void drift, flow pattern distribution, and stability, a scaled loop geometry is designed. For modeling the steam/water flow in a BWR fuel assembly, scaling criteria are derived using the one-dimensional drift-flux model. Thermal equilibrium and subcooled boiling conditions are treated separately, resulting in one overall set of criteria. Scaling on all flow regimes that can be present in a normal fuel assembly leads to fixing both the assembly mass flux and the geometric dimensions. When Freon-12 is used as a modeling fluid, model assembly dimensions must be 0.46 of the prototype. Total power consumption must be reduced by a factor 50. To sustain cooling by natural circulation, a modeled chimney and downcomer are included

  12. 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. In this study, the sodium experiments were conducted using a sodium test loop PLANDTL in order to investigate the effect of operation mode on transient behavior of thermal hydraulic in PRACS loop. The experimental results revealed the effect of increasing heat removal capacity of PRACS and the forced flow operation in PRACS loop on the thermal transient in the PRACS loop and natural circulation behavior of PRACS. (author)

  13. Engineered safety feature, an emergency core cooling system at Pakistan research reactor-1

    International Nuclear Information System (INIS)

    In the present study effectiveness of emergency core cooling system (ECCS) has been studied in case loss of coolant accident occurs at Pakistan research reactor (PARR-1). The reactor is a swimming pool type using MTR fuel. It was converted from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel in 1992. It was also upgraded from a steady-state power level of 5-10 MW. Several additional facilities were provided to satisfy the requirements of enhanced power level. For safety consideration, emergency core cooling system (ECCS) was also installed to avoid any possibility of core meltdown. Evaluation of ECCS has been carried out for which standard correlations have been employed to find peak clad temperature profile after loss of coolant accident

  14. Engineered safety feature, an emergency core cooling system at Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, Ishtiaq Hussain [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)], E-mail: ishtiaq@pinstech.org.pk; Mahmood, Tariq [Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad (Pakistan)

    2008-06-15

    In the present study effectiveness of emergency core cooling system (ECCS) has been studied in case loss of coolant accident occurs at Pakistan research reactor (PARR-1). The reactor is a swimming pool type using MTR fuel. It was converted from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel in 1992. It was also upgraded from a steady-state power level of 5-10 MW. Several additional facilities were provided to satisfy the requirements of enhanced power level. For safety consideration, emergency core cooling system (ECCS) was also installed to avoid any possibility of core meltdown. Evaluation of ECCS has been carried out for which standard correlations have been employed to find peak clad temperature profile after loss of coolant accident.

  15. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  16. Comparative analyses on nuclear characteristics of water-cooled breeder cores

    International Nuclear Information System (INIS)

    In order to compare the nuclear characteristics of water-cooled breeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. First, the comparison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and η-value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. In order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1 keV). And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of water-cooled

  17. Cool Core Bias in Sunyaev-Zel'dovich Galaxy Cluster Surveys

    CERN Document Server

    Lin, Henry W; Benson, Bradford; Miller, Eric

    2015-01-01

    Sunyaev-Zeldovich (SZ) surveys find massive clusters of galaxies by measuring the inverse Compton scattering of cosmic microwave background off of intra-cluster gas. The cluster selection function from such surveys is expected to be nearly independent of redshift and cluster astrophysics. In this work, we estimate the effect on the observed SZ signal of centrally-peaked gas density profiles (cool cores) and radio emission from the brightest cluster galaxy (BCG) by creating mock observations of a sample of clusters that span the observed range of classical cooling rates and radio luminosities. For each cluster, we make simulated SZ observations by the South Pole Telescope and characterize the cluster selection function, but note that our results are broadly applicable to other SZ surveys. We find that the inclusion of a cool core can cause a change in the measured SPT significance of a cluster between 0.01% - 10% at z > 0.3, increasing with cuspiness of the cool core and angular size on the sky of the cluster ...

  18. 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.

  19. The effect of lower body cooling on the changes in three core temperature indices

    International Nuclear Information System (INIS)

    Rectal (Tre), ear canal (Tear) and esophageal (Tes) temperatures have been used in the literature as core temperature indices in humans. The aim of the study was to investigate if localized lower body cooling would have a different effect on each of these measurements. We hypothesized that prolonged lower body surface cooling will result in a localized cooling effect for the rectal temperature not reflected in the other core measurement sites. Twelve participants (mean ± SD; 26.8 ± 6.0 years; 82.6 ± 13.9 kg; 179 ± 10 cm, BSA = 2.00 ± 0.21 m2) attended one experimental session consisting of sitting on a rubberized raft floor surface suspended in 5 °C water in a thermoneutral air environment (∼21.5 ± 0.5 °C). Experimental conditions were (a) a baseline phase during which participants were seated for 15 min in an upright position on an insulated pad (1.408 K . m2 . W−1); (b) a cooling phase during which participants were exposed to the cooling surface for 2 h, and (c) an insulation phase during which the baseline condition was repeated for 1 h. Temperature data were collected at 1 Hz, reduced to 1 min averages, and transformed from absolute values to a change in temperature from baseline (15 min average). Metabolic data were collected breath-by-breath and integrated over the same temperature epoch. Within the baseline phase no significant change was found between the three indices of core temperature. By the end of the cooling phase, Tre was significantly lower (Δ = −1.0 ± 0.4 °C) from baseline values than from Tear (Δ = −0.3 ± 0.3 °C) and Tes (Δ = −0.1 ± 0.3 °C). Tre continued to decrease during the insulation phase from Δ −1.0 ± 0.4 °C to as low as Δ −1.4 ± 0.5 °C. By the end of the insulation phase Tre had slightly risen back to Δ −1.3 ± 0.4 °C but remained significantly different from baseline values and from the other two core measures. Metabolic data showed no variation throughout the experiment. In conclusion, the local

  20. 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.

  1. Analysis on Roles for Components of Passive Emergency Core Cooling System

    International Nuclear Information System (INIS)

    International nuclear industry has been adopting a passive safety system to enhance safety and reliability of nuclear power plant with an advanced technology. Also, domestic nuclear industry issued the necessity for the development of key technologies for passive safety system design. It is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a Passive Emergency Core Cooling System (PECCS) is to be adopted as an improved safety design feature of APR+. When unfavorable accidents such as Station Black Out(SBO) happen, the PECCS should be able to make up the core and then cool down the core. This study discusses the applicability of PECCS and the proper design combinations especially during SBO. In this study, the applicability of PECCS and analysis on roles of components during SBO were assessed. RELAP5 calculations show that PECCS can make up the core and then prevent the core from being damaged during SBO with PAFS unavailable. Resultant analysis shows the role of the ADV for RCS depressurization, and SITs for RCS making up. When PAFS is available, ADVs is not required. Further study is required to sensitivity analysis such as actuation signal and setpoint

  2. Analysis on Roles for Components of Passive Emergency Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Soon Il; Hong, Soon Joon [FNC Tech, Yongin (Korea, Republic of); Kang, Sang Hee; Kim, Han Gon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    International nuclear industry has been adopting a passive safety system to enhance safety and reliability of nuclear power plant with an advanced technology. Also, domestic nuclear industry issued the necessity for the development of key technologies for passive safety system design. It is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a Passive Emergency Core Cooling System (PECCS) is to be adopted as an improved safety design feature of APR+. When unfavorable accidents such as Station Black Out(SBO) happen, the PECCS should be able to make up the core and then cool down the core. This study discusses the applicability of PECCS and the proper design combinations especially during SBO. In this study, the applicability of PECCS and analysis on roles of components during SBO were assessed. RELAP5 calculations show that PECCS can make up the core and then prevent the core from being damaged during SBO with PAFS unavailable. Resultant analysis shows the role of the ADV for RCS depressurization, and SITs for RCS making up. When PAFS is available, ADVs is not required. Further study is required to sensitivity analysis such as actuation signal and setpoint.

  3. A simple model of cooling neutron stars with superfluid core comparison with observations

    CERN Document Server

    Levenfish, K P; Yakovlev, D G; Shibanov, Yu.A.

    1999-01-01

    Cooling of neutron stars (NSs) with superfluid cores is simulated taking into account neutrino emission produced by Cooper pairing of nucleons. The critical temperatures of neutron and proton superfluidities, $T_{cn}$ and $T_{cp}$, are assumed to be constant over the NS core, and treated as free parameters. They are constrained using the surface temperatures $T_s$ of isolated NSs (RX J0822-43, PSR 1055-52, 1E 1207-52, Vela, Geminga, PSR 0656+14, RX J0002+62), obtained by interpretation of observed thermal radiation either with black body spectrum or with hydrogen atmosphere models.

  4. Deep high-resolution X-ray spectra from cool-core clusters

    OpenAIRE

    Sanders, J.S.; Fabian, A.C.; Frank, K. A.; Peterson, J. R.; Russell, H. R.

    2009-01-01

    We examine deep XMM-Newton Reflection Grating Spectrometer (RGS) spectra from the cores of three X-ray bright cool core galaxy clusters, Abell 262, Abell 3581 and HCG 62. Each of the RGS spectra show Fe XVII emission lines indicating the presence of gas around 0.5 keV. There is no evidence for O VII emission which would imply gas at still cooler temperatures. The range in detected gas temperature in these objects is a factor of 3.7, 5.6 and 2 for Abell 262, Abell 3581 and HCG 62, respectively...

  5. Fast reactor core thermal-hydraulic analyses during transition from forced to natural circulation

    International Nuclear Information System (INIS)

    The modeling for inter-subchannel mixing effects was presented to simulate the fast reactor transition from rated to natural circulation decay heat removal conditions which was usually accompanied by all flow regimes: forced, mixed and natural convection. The model was constructed based on correlations for mixing and pressure drop coefficients developed at MIT. This correlation was originally proposed for steady states subchannel analyses. In the present study, application of the mixing correlation was extended to unsteady multi-dimensional analyses by introducing a threshold function. The function enabled to switch the correlations adequately with change of the flow regimes, depending on Richardson number which is an index of buoyancy effect on the flow field. The modeling was validated through calculation of sodium experiments featuring 37, 61 and 169-pin bundle subassemblies. Comparisons of the experimental and numerical results revealed that the modeling was capable of predicting the core thermal-hydraulic field under wide spectrum of flow rate and heating conditions. (author)

  6. Hydraulic analysis of the emergency core cooling system of the RP-10 reactor

    International Nuclear Information System (INIS)

    This work shows calculation for the hydraulic analysis of the Emergency Core Cooling System (ECCS) of the RP-10 Reactor. This analysis is necessary for the design of such system. According to calculation results shown in the graphics, a pipe line of two inches of nominal diameter should be selected for such system and a maximum flow of 5 m3/h should be reached

  7. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    International Nuclear Information System (INIS)

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MWth) and of its demonstrator reactor (300 MWth) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors

  8. Improving the reliability modeling concerning the emergency core cooling system at Gentilly-2 Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Komljenovic, D.; Vaillancourt, R.; Croteau, M. [Hydro Quebec, Gentilly-2, Nuclear Generating Station, Quebec (Canada); Abdul-Nour, G.; Darragi, M. [Univ. du Quebec a Trois Rivieres, Trois Rivieres, Quebec (Canada)

    2003-07-01

    This technical paper presents an approach to improving the reliability modeling concerning the performance of the Emergency Core Cooling (ECC) System at Gentilly-2 Nuclear Generating Station following a loss of coolant accident (LOCA). It includes a quantitative unavailability analysis based on the current system design and operation. The study is performed as a part of a project with regard to an extension of the plant planned outage period. (author)

  9. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grasso, G., E-mail: giacomo.grasso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Petrovich, C., E-mail: carlo.petrovich@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Mattioli, D., E-mail: davide.mattioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Artioli, C., E-mail: carlo.artioli@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Sciora, P., E-mail: pierre.sciora@cea.fr [CEA (Alternative Energies and Atomic Energy Commission), DEN, DER, 13108 St Paul lez Durance (France); Gugiu, D., E-mail: daniela.gugiu@nuclear.ro [RATEN-ICN (Institute for Nuclear Research), Cod 115400 Mioveni, Str. Campului, 1, Jud. Arges (Romania); Bandini, G., E-mail: giacomino.bandini@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), via Martiri di Monte Sole, 4, 40129 Bologna (Italy); Bubelis, E., E-mail: evaldas.bubelis@kit.edu [KIT (Karlsruhe Institute of Technology), Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K., E-mail: konstantin.mikityuk@psi.ch [PSI (Paul Scherrer Institute), OHSA/D11, 5232 Villigen PSI (Switzerland)

    2014-10-15

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MW{sub th}) and of its demonstrator reactor (300 MW{sub th}) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors.

  10. Chandra Observation of the Interaction of the Radio Source and Cooling Core in Abell 2063

    CERN Document Server

    Kanov, K N; Hicks, A K; Kanov, Kalin N.; Sarazin, Craig L.; Hicks, Amalia K.

    2006-01-01

    We present the results of a Chandra observation of the cooling core cluster Abell 2063. Spectral analysis shows that there is cool gas (2 keV) associated with the cluster core, which is more than a factor of 2 cooler than the outer cluster gas (4.1 keV). There also is spectral evidence for a weak cooling flow, Mdot ~ 20 Msun/yr. The cluster exhibits a complex structure in the center that consists of several bright knots of emission, a depression in the emission to the north of the center of the cluster, and a shell of emission surrounding it. The depression in the X-ray emission is coincident with the position of the north-eastern radio lobe of the radio source associated with the cluster-central galaxy. The shell surrounding this region appears to be hotter, which may be the result of a shock that has been driven into the gas by the radio source. The power output of the radio source appears to be sufficient to offset the cooling flow, and heating of the gas through shocks is a possible explanation of how the...

  11. Sodium-cooled Fast Reactor Core Designs for the TRU burning with Thorium blanket

    International Nuclear Information System (INIS)

    In this study, the SFR(Sodium-cooled Fast Reactor) burner cores are designed with thorium blanket to have smaller burnup reactivity swing but higher TRU burning capability than the typical SFR burner cores using the TRU-U-10Zr fuel. Furthermore, we expect the SFR burner cores using thorium blanket have smaller coolant void reactivity because of the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. From the results, it is found that use of the thorium blanket both in inner and outer cores gives several desirable features such as the reduction of sodium void worth, small burnup reactivity swing but less negative Doppler coefficient and reduced control rod worth and that the use of thorium blanket only in the inner core gives much smaller sodium void worth but larger burnup reactivity swing than the cores using thorium blanket both in the inner and outer cores

  12. Nonlinear dynamic analysis of prismatic elements for high-temperature gas-cooled reactor cores

    International Nuclear Information System (INIS)

    The high-temperature gas-cooled reactor (HTGR) core consists of several thousand prismatic graphite fuel elements arranged in columns within a prestressed concrete vessel. A major research and development effort was initiated in 1970 at General Atomic Company to study the dynamic response of the HTGR core arrangement to seismic excitation. A discussion is pesented of the history and some of the results of this effort with respect to the advances made in the development of analytical methods. The computer programs developed to perform the analysis are described, along with certain techniques and the modeling required to utilize them. The nonlinear dynamic analysis techniques employed to analyze the HTGR core are described

  13. A 3.55 keV line from DM → a → γ: predictions for cool-core and non-cool-core clusters

    Energy Technology Data Exchange (ETDEWEB)

    Conlon, Joseph P.; Powell, Andrew J., E-mail: j.conlon1@physics.ox.ac.uk, E-mail: andrew.powell2@physics.ox.ac.uk [Rudolf Peierls Centre for Theoretical Physics, University of Oxford, 1 Keble Road, Oxford, OX1 3NP (United Kingdom)

    2015-01-01

    We further study a scenario in which a 3.55 keV X-ray line arises from decay of dark matter to an axion-like particle (ALP), that subsequently converts to a photon in astrophysical magnetic fields. We perform numerical simulations of Gaussian random magnetic fields with radial scaling of the magnetic field magnitude with the electron density, for both cool-core 'Perseus' and non-cool-core 'Coma' electron density profiles. Using these, we quantitatively study the resulting signal strength and morphology for cool-core and non-cool-core clusters. Our study includes the effects of fields of view that cover only the central part of the cluster, the effects of offset pointings on the radial decline of signal strength and the effects of dividing clusters into annuli. We find good agreement with current data and make predictions for future analyses and observations.

  14. Application of MF,Ozone and RO in Treatment of Municipal Sewage Reused as Circulating Cooling Water

    Institute of Scientific and Technical Information of China (English)

    Zhang Liqiang

    2007-01-01

    @@ Reuse of treated municipal sewage as circulating cooling water of fossil-fired power plants is a very theme worthy to be studied and spread because of the water shortage in most areas of China. This paper presents a process using coagulation + MF + ozone + partial RO to deal with the recycled sewage after treated preliminarily in sewage treatment plant. The process solves effectively the problem of higher TDS and higher total hardness in product water in winter, thus is especially fit for cities where sewage quality changes obviously with seasons.

  15. Radio mini-halos and AGN heating in cool core clusters of galaxies

    CERN Document Server

    Gitti, Myriam

    2016-01-01

    The brightest cluster galaxy (BCG) in the majority of relaxed, cool core galaxy clusters is radio loud, showing non-thermal radio jets and lobes ejected by the central active galactic nucleus (AGN). Such relativistic plasma has been unambiguously shown to interact with the surrounding thermal intra-cluster medium (ICM) thanks to spectacular images where the lobe radio emission is observed to fill the cavities in the X-ray-emitting gas. This `radio-mode AGN feedback' phenomenon, which is thought to quench cooling flows, is widespread and is critical to understand the physics of the inner regions of galaxy clusters and the properties of the central BCG. At the same time, mechanically-powerful AGN are likely to drive turbulence in the central ICM which may contribute to gas heating and also play a role for the origin of non-thermal emission on cluster-scales. Diffuse non-thermal emission has been observed in a number of cool core clusters in the form of a radio mini-halo surrounding the radio-loud BCG on scales ...

  16. Performance studies of a new core cooling monitor in a boiling water reactor

    International Nuclear Information System (INIS)

    Performance studies of a new type of core cooling monitors have been carried out in the Barsebaeck Nuclear Power Station during the operation periods 1988-10-04 to 1989-07-05, 1989-08-03 to 1990-09-05 and 1990-09-28 to 1991-07-04. The results showed that the monitors, which were placed inside the reactor core, are very sensitive to variations of the reactor operating conditions, and that 34 months of irradiation did not influence the signals from the monitors. Experiments were also carried out in a 160 bar loop, where sudden uncovers of the monitors were achieved by decreasing the liquid level of the coolant surrounding the monitors. The experiments included the pressures of 5, 20, 50, 70 and 155 bar, and the responses to uncover were in the ranges between 11 and 82 mV/sec or a total step change of 2 V at typical BWR conditions. This is of the order of two decades higher than the responses from monitors based on thermocouple readings. The monitors can be operated in two modes, the core cooling mode and the temperature mode. In the former mode the electrical current is 3-4 A, and in the latter mode, where the monitor actually serves as a thermometer, the current is in the order of 50-100 mA. In the laboratory the monitors have been studied for temperatures up to 1265 deg. C, which is very useful in case of a severe reactor accident. Thus, during such events the temperatures in the reactor core could be followed up to this level and the monitors could also be used to activate certain safety equipment. The function as well as the design of the instrument is verified in laboratory experiments, computer calculations and reactor tests and is now ready for implementation in the BWR instrumentation. In summary: 1. The proposed monitor can operate in two modes; the core cooling mode and the temperature mode. 2. Laboratory studies have shown that the responses to uncover are two decades higher than signals from monitors based on thermocouple readings. 3. No effects of

  17. Utilization of control rod drive (CRD) system for long term core cooling

    International Nuclear Information System (INIS)

    In this paper we consider an application of Probabilistic Risk Assessment (PRA) to risk management. Foreseeable risk management strategies to prevent core damage are constrained by the availability of first line systems as well as support systems. The actual trend in the evaluation of risk management options can be performed in a number of ways. An example is the identification of back-up systems which could be used to perform the same safety functions. In this work we deal with the evaluation of the feasibility, for BWR's, to use the Control Rod Drive system to maintain an adequate reactor core long term cooling in some accident sequences. This preliminary evaluation is carried out as a part of the Internal Events Analysis for Laguna Verde Nuclear Power Plant (LVNPP) that is currently under way by the Mexican Nuclear Regulatory Body. This analysis addresses the evaluation and incorporation of all the systems, including the safety related and the back-up non safety related systems, that are available for the operator in order to prevent core damage. As a part of this analysis the containment venting capability is also evaluated as a back-up of the containment heat removal function. This will prevent the primary containment overpressurization and loss of certain core cooling systems. A selection of accident sequences in which the Control Rod Drive system could be used to mitigate the accident and prevent core damage are discussed. A personal computer transient analysis code is used to carry out thermohydraulic simulations in order to evaluate the Control Rod Drive system performance, the corresponding results are presented. Finally, some preliminary conclusions are drawn. (author). 9 refs, 5 figs

  18. Passive Core Cooling Systems for Next Generation NPPs: Characteristics and State of the Art

    Energy Technology Data Exchange (ETDEWEB)

    Morozov, Andrey; Soshkina, Alexandra [Institute for Physics and Power Engineering by A.I. Leypunsky, 1 Bondarenko sq. Obninsk, 249033 (Russian Federation)

    2008-07-01

    Among nuclear power generation plants, light water reactors are mainly used at present, and are anticipated to be predominant in the future. To improve the light water reactors the development of the LWRs for the next generation is carried out at various organizations. For example, in the USA the Westinghouse AP-1000 is based on proven technology but with an emphasis on passive safety features. The reactor passive core cooling systems include the core makeup tanks system, passive residual heat removal heat exchanger and in-containment refuelling water storage tank. In Russia has been developed the so-called NPP-2006 project of a VVER-1200 nuclear power plant with a V-392M reactor unit. To provide the safety, protection passive systems which do not depend upon human errors are widely used in this project. Among these are hydro-tanks of the second stage and passive heat removal system. In the presented paper an overview of passive core cooling systems for next generation NPPs is given. (authors)

  19. Radio and Deep Chandra Observations of the Disturbed Cool Core Cluster Abell 133

    CERN Document Server

    Randall, S W; Nulsen, P E J; Owers, M S; Sarazin, C L; Forman, W R; Murray, S S

    2010-01-01

    We present results based on new Chandra and multi-frequency radio observations of the disturbed cool core cluster Abell 133. The diffuse gas has a complex bird-like morphology, with a plume of emission extending from two symmetric wing-like features, and capped with a filamentary radio relic. X-ray observations indicate the presence of either high temperature gas or non-thermal emission in the region of the relic. We find evidence for a weak elliptical X-ray surface brightness edge surrounding the core, consistent with a sloshing cold front. The plume is consistent with having formed due to uplift by a buoyantly rising radio bubble, now seen as the radio relic. Our results are inconsistent with the previous suggestion that the X-ray wings formed due to the passage of a weak shock through the cool core. We instead conclude that the wings are due to X-ray cavities formed by gas displacement by the radio relic. The central cD galaxy contains two small-scale cold gas clumps that are slightly offset from their opt...

  20. Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

    International Nuclear Information System (INIS)

    Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

  1. Advanced sodium cooled reactor cores having thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    In this paper, a design concept of 400 MWe sodium cooled fast reactor (SFR) cores having thorium blankets for effective burning of TRU (Transuranics) from LWR spent fuel is described. Specifically, we considered two recycling options of thorium blankets : 1) no recycling and 2) fully recycling. The thorium blankets are loaded in the axially central regions of the core regions and their axial heights are adjusted so as to increase TRU burning rate and to reduce burnup reactivity swing. Also, we analyzed the performances of the cores having different fuel management batch sizes and different recycling options for the searched core configuration. The results show that the axial thorium blankets with no recycling option can be effectively used to increase TRU burning rate with a significant reduction of burnup reactivity swing in comparison with typical SFR burner cores having no blankets while the recycling of thorium blanket degrades TRU burning rate and burnup reactivity swing but it leads to a reduction of sodium void worth and more negative Doppler coefficient. (author)

  2. Deep Chandra study of the truncated cool core of the Ophiuchus cluster

    CERN Document Server

    Werner, N; Canning, R E A; Allen, S W; King, A L; Sanders, J S; Simionescu, A; Taylor, G B; Morris, R G; Fabian, A C

    2016-01-01

    We present the results of a deep (280 ks) Chandra observation of the Ophiuchus cluster, the second-brightest galaxy cluster in the X-ray sky. The cluster hosts a truncated cool core, with a temperature increasing from kT~1 keV in the core to kT~9 keV at r~30 kpc. Beyond r~30 kpc the intra-cluster medium (ICM) appears remarkably isothermal. The core is dynamically disturbed with multiple sloshing induced cold fronts, with indications for both Rayleigh-Taylor and Kelvin-Helmholtz instabilities. The sloshing is the result of the strongly perturbed gravitational potential in the cluster core, with the central brightest cluster galaxy (BCG) being displaced southward from the global center of mass. The residual image reveals a likely subcluster south of the core at the projected distance of r~280 kpc. The cluster also harbors a likely radio phoenix, a source revived by adiabatic compression by gas motions in the ICM. Even though the Ophiuchus cluster is strongly dynamically active, the amplitude of density fluctuat...

  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. Keeping mush mushy at the core-mantle boundary: the role of internal convection and secular cooling

    Science.gov (United States)

    Hernlund, J. W.; Jellinek, M.

    2007-12-01

    Williams and Garnero (1996) proposed that thin (5-40 km thick) patches of dramatically decreased seismic velocity above the core-mantle boundary (ultralow-velocity zones, or ULVZ) could best be explained by the presence of partial melt, and this remains the favored mechanism to explain the anomalous seismic properties. However, simple estimates for compaction and expulsion of melt from a porous solid on the order of 1 Gyr require an effective bulk viscosity that is probably much larger than realistic, and therefore melt should have separated long ago from the interstices it occupies in the matrix. In more detail, however, this layer is subject to a more complicated style of internal stirring governed by the combined influences of motions induced by flow in the overlying mantle and motions arising in response to the compaction-driven drainage of interstitial melt, which depend critically on the melt fraction. In the simplest scenario, analogous to the sedimentation of solids from a convecting slurry, this circulation may enhance expulsion of fluid from the mush. Additional important factors include the possibility of melting and freezing in different parts of the layer due, for example, to small thermal gradients, slow secular cooling at the top of the core, chemical flux to or from the core, and internal compositional stratification. We use numerical models of compaction and flow in a two-phase medium in equilibrium according to a simple binary phase diagram to better understand the evolution of porosity in a churning mush. Flow is driven by convection in the overlying mantle along with internal buoyancy forces due to phase and composition variations, with the former becoming more important when the tendency is toward a gravitationally stable stratification of the mush. Flux of light elements to or from the core is also studied by imposing composition at the lower boundary. A central aim of this work is to identify plausible sets of conditions in which thin

  5. Natural circulation data and methods for advanced water cooled nuclear power plant designs. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The complex set of physical phenomena that occur in a gravity environment when a geometrically distinct heat sink and heat source are connected by a fluid flow path can be identified as natural circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established. Within the present context, natural convection is used to identify the phenomena that occur when a heat source is put in contact with a fluid. Therefore, natural convection characterizes a heat transfer regime that constitutes a subset of NC phenomena. This report provides the presented papers and summarizes the discussions at an IAEA Technical Committee Meeting (TCM) on Natural Circulation Data and Methods for innovative Nuclear Power Plant Design. While the planned scope of the TCM involved all types of reactor designs (light water reactors, heavy water reactors, gas-cooled reactors and liquid metal-cooled reactors), the meeting participants and papers addressed only light water reactors (LWRs) and heavy water reactors (HWRs). Furthermore, the papers and discussion addressed both evolutionary and innovative water cooled reactors, as defined by the IAEA. The accomplishment of the objectives of achieving a high safety level and reducing the cost through the reliance on NC mechanisms, requires a thorough understanding of those mechanisms. Natural circulation systems are usually characterized by smaller driving forces with respect to the systems that use an external source of energy for the fluid motion. For instance, pressure drops caused by vertical bends and siphons in a given piping system, or heat losses to environment are a secondary design consideration when a pump is installed and drives the flow. On the contrary, a significant influence upon the overall system performance may be expected due to the same pressure drops and thermal power release to the environment when natural circulation produces the coolant flow. Therefore, the level of knowledge for

  6. The effects of aging on Boiling Water Reactor core isolation cooling system

    International Nuclear Information System (INIS)

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes

  7. Fuel performance models for high-temperature gas-cooled reactor core design

    International Nuclear Information System (INIS)

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10-4 fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience

  8. 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

  9. Severe accident assessment: development of the gas flux dryout model for cooling of core debris

    International Nuclear Information System (INIS)

    A model for boiling and dryout in a particle debris bed with permeable boundary conditions is developed and compared with various dryout models, and incorporated into the modified MARCH/KAERI computer code to analyze for the combined mechanisms of thermal interactions. Comparative and parametric studies show that the particle sizes have an important effect on debris bed cooling but not apparent effect on the magnitude of peak pressure in the containment building. It is also shown that the gas flux model represents an improvement of the combined thermal interactions among core debris, water and gas over the previous models. (Author)

  10. Uncertainty calculation of emergency core cooling system for boiling water reactor (BWR-5)

    International Nuclear Information System (INIS)

    A brief description about uncertainty calculation of emergency core cooling system for boiling water reactor (BWR-5) is presented in this paper. Based on methodology of PSA level 1 and draft description of ECCS's document supplied by TOSHIBA (Code PSO-00-00097, July 2000) the event tree is built. The fault trees of three of subsystems HPCSS, LPCSS, LPCIS can be developed due to the simplified P and ID of ECCS and the reliability data accompanied. The computer code used to develop fault tree is KIRAP-tree and one used to find cut set and calculated uncertainty is KCUT. (author)

  11. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

  12. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities

  13. Ram pressure stripping of the cool core of the Ophiuchus Cluster

    CERN Document Server

    Million, E T; Werner, N; Taylor, G B

    2009-01-01

    We report results from a Chandra study of the central regions of the nearby, X-ray bright, Ophiuchus Cluster (z = 0.03), the second-brightest cluster in the sky. Our study reveals a dramatic, close-up view of the sloshing, stripping and potential destruction of a cool core within a rich cluster. The X-ray emission from the Ophiuchus Cluster core exhibits a comet-like morphology extending to the north, driven by merging activity, indicative of ram-pressure stripping caused by rapid motion through the ambient cluster gas. A cold front at the southern edge implies a velocity of 1000$\\pm$200 km/s (M~0.6). The X-ray emission from the cluster core is sharply peaked. However, the peak is offset by 4 arcsec (~2 kpc) from the optical center of the associated cD galaxy. This indicates that ram pressure has slowed the core, allowing the relatively collisionless stars and dark matter to carry on ahead. The cluster exhibits the strongest central temperature gradient of any massive cluster observed to date: the temperature...

  14. Chandra Observations of the Disruption of the Cool Core in Abell 133

    CERN Document Server

    Fujita, Y; Kempner, J C; Rudnick, L; Roy, L; Andernach, H; Ehle, M; Slee, A; Fujita, Yutaka; Sarazin, Craig L.; Kempner, Joshua C.

    2002-01-01

    We present the analysis of a Chandra observation of the galaxy cluster Abell 133, which has a cooling flow core, a central radio source, and a diffuse, filamentary radio source which has been classified as a radio relic. The X-ray image shows that the core has a complex structure. The most prominent feature is a "tongue" of emission which extends from the central cD galaxy to the northwest and partly overlaps the radio relic. One possibility is that this tongue is produced by Kelvin-Helmholtz (KH) instabilities through the interaction between the cold gas around the cD galaxy and hot intracluster medium. We estimate the critical velocity and time scale for the KH instability to be effective for the cold core around the cD galaxy. We find that the KH instability can disrupt the cold core if the relative velocity is >~400 km s^-1. We compare the results with those of clusters in which sharp, undisrupted cold fronts have been observed; in these clusters, the low temperature gas in their central regions has a mor...

  15. A particle-bed gas-cooled fast reactor core design for waste minimization

    International Nuclear Information System (INIS)

    The issue of waste minimization in advanced reactor systems has been investigated using the Particle-Bed Gas-cooled Fast Reactor (PB-GCFR) design being developed and funded under the U.S. Department of Energy Nuclear Energy Research Initiative (USDOE NERI) Programme. Results indicate that for the given core power density and constraint on the maximum TRU enrichment allowable, the lowest amount of radio-toxic transuranics to be processed and hence sent to the repository is obtained for long-life core designs. Calculations were additionally done to investigate long-life core designs using LWR spent fuel TRU and recycle TRU, and different feed, matrix and reflector materials. The recycled TRU and LWR spent TRU fuels give similar core behaviours, because of the fast spectrum environment which does not significantly degrade the TRU composition. Using light elements as reflector material was found to be unattractive because of power peaking problems and large reactivity swings. The application of a lead reflector gave the longest cycle length and lowest TRU processing requirement. Materials compatibility and performance issues require additional investigation. (author)

  16. An improved method for calculating control rod reactivity worths in fast sodium cooled reactor cores

    International Nuclear Information System (INIS)

    An improved method is presented to determine the reactivities of strongly inhomogeneous control rod arrangements in fast sodium cooled reactor cores. The method is based on a detailed evaluation of the multiplication constants for the rods embedded in a large surrounding of fuel material. These calculations are performed using two-dimensional transport theory, with an accurate representation of the actual geometry in RΘ coordinates and with fine discretizations in coordinate space and energy. Three-dimensional whole core calculations are carried out in diffusion approximation, with a coarse spatial hexagonal-Z mesh and few energy groups, replacing the individual reactor cells by homogeneous arrangements. The homogenized macroscopic group cross sections are generated with standard methods, however using reduced boron contents of the absorber pins as compared with their actual values. The appropriate boron concentrations are found by comparing the control rod reactivity worths resulting from the two-dimensional transport calculations with those determined from corresponding diffusion calculations with homogenized compositions for the corresponding regions, which possess as many features of the final whole core calculations as possible. In this way, the corrections necessitated by the heterogeneity, transport, mesh, and condensation effects are incorporated in the macroscopic cross sections. With these as input, the computed rod worths of the secondary shutdown system of the SUPERPHENIX-1 (SPX-1) power production core are essentially improved as compared with results of earlier calculations. This progress of the calculational method is clearly demonstrated by a comparison with measured reactivity worths. (orig.)

  17. Comparative Analysis of Effectiveness of Various Emergency Core Cooling System Design Options for Sodium Fast Reactors of High Rower

    International Nuclear Information System (INIS)

    Effectiveness of various design options for emergency core cooling systems has been compared as applied to a pool type sodium fast reactor of high power. Thermal hydraulic parameters of the reactor under cooling conditions are analyzed with the use of the Russian thermal hydraulic code GRIF which allows 3D velocity and temperature fields to be calculated in the reactor, with account of thermal hydraulic processes in the core inter-wrapper space. To realize the cooling system margin in case of additional parallel failures, the parameters were calculated for the cooling mode accompanied with additional conditions of malfunction of part of emergency heat exchangers (DHX) for unknown reasons. Based one the calculation analysis the conclusion is made about a relative effectiveness of the emergency cooling system design options considered. (author)

  18. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  19. Chandra Observation of Abell 1142: A Cool-Core Cluster Lacking a Central Brightest Cluster Galaxy?

    CERN Document Server

    Su, Yuanyuan; Gastaldello, Fabio; van Weeren, Reinout

    2016-01-01

    Abell~1142 is a low-mass galaxy cluster at low redshift containing two comparable Brightest Cluster Galaxies (BCG) resembling a scaled-down version of the Coma Cluster. Our Chandra analysis reveals an X-ray emission peak, roughly 100 kpc away from either BCG, which we identify as the cluster center. The emission center manifests itself as a second beta-model surface brightness component distinct from that of the cluster on larger scales. The center is also substantially cooler and more metal rich than the surrounding intracluster medium (ICM), which makes Abell 1142 appear to be a cool core cluster. The redshift distribution of its member galaxies indicates that Abell 1142 may contain two subclusters with each containing one BCG. The BCGs are merging at a relative velocity of ~1200 km/s. This ongoing merger may have shock-heated the ICM from ~ 2 keV to above 3 keV, which would explain the anomalous L_X--T_X scaling relation for this system. This merger may have displaced the metal-enriched "cool core" of eith...

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

    International Nuclear Information System (INIS)

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO2) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D.; Sterbentz, J. [Idaho National Engineering and Environmental Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States); Meyer, M. [Argonne National Laboratory- West (United States); Lowden, R. [Oak Ridge National Laboratory (United States); Hoffman, E.; Wei, T.Y.C. [Argonne National Laboratory (United States)]. e-mail: weavkd@inel.gov

    2004-07-01

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

  2. North and equatorial Pacific Ocean circulation in the CORE-II hindcast simulations

    Science.gov (United States)

    Tseng, Yu-heng; Lin, Hongyang; Chen, Han-ching; Thompson, Keith; Bentsen, Mats; Böning, Claus W.; Bozec, Alexandra; Cassou, Christophe; Chassignet, Eric; Chow, Chun Hoe; Danabasoglu, Gokhan; Danilov, Sergey; Farneti, Riccardo; Fogli, Pier Giuseppe; Fujii, Yosuke; Griffies, Stephen M.; Ilicak, Mehmet; Jung, Thomas; Masina, Simona; Navarra, Antonio; Patara, Lavinia; Samuels, Bonita L.; Scheinert, Markus; Sidorenko, Dmitry; Sui, Chung-Hsiung; Tsujino, Hiroyuki; Valcke, Sophie; Voldoire, Aurore; Wang, Qiang; Yeager, Steve G.

    2016-08-01

    We evaluate the mean circulation patterns, water mass distributions, and tropical dynamics of the North and Equatorial Pacific Ocean based on a suite of global ocean-sea ice simulations driven by the CORE-II atmospheric forcing from 1963-2007. The first three moments (mean, standard deviation and skewness) of sea surface height and surface temperature variability are assessed against observations. Large discrepancies are found in the variance and skewness of sea surface height and in the skewness of sea surface temperature. Comparing with the observation, most models underestimate the Kuroshio transport in the Asian Marginal seas due to the missing influence of the unresolved western boundary current and meso-scale eddies. In terms of the Mixed Layer Depths (MLDs) in the North Pacific, the two observed maxima associated with Subtropical Mode Water and Central Mode Water formation coalesce into a large pool of deep MLDs in all participating models, but another local maximum associated with the formation of Eastern Subtropical Mode Water can be found in all models with different magnitudes. The main model bias of deep MLDs results from excessive Subtropical Mode Water formation due to inaccurate representation of the Kuroshio separation and of the associated excessively warm and salty Kuroshio water. Further water mass analysis shows that the North Pacific Intermediate Water can penetrate southward in most models, but its distribution greatly varies among models depending not only on grid resolution and vertical coordinate but also on the model dynamics. All simulations show overall similar large scale tropical current system, but with differences in the structures of the Equatorial Undercurrent. We also confirm the key role of the meridional gradient of the wind stress curl in driving the equatorial transport, leading to a generally weak North Equatorial Counter Current in all models due to inaccurate CORE-II equatorial wind fields. Most models show a larger

  3. REMOVING COOL CORES AND CENTRAL METALLICITY PEAKS IN GALAXY CLUSTERS WITH POWERFUL ACTIVE GALACTIC NUCLEUS OUTBURSTS

    International Nuclear Information System (INIS)

    Recent X-ray observations of galaxy clusters suggest that cluster populations are bimodally distributed according to central gas entropy and are separated into two distinct classes: cool core (CC) and non-cool core (NCC) clusters. While it is widely accepted that active galactic nucleus (AGN) feedback plays a key role in offsetting radiative losses and maintaining many clusters in the CC state, the origin of NCC clusters is much less clear. At the same time, a handful of extremely powerful AGN outbursts have recently been detected in clusters, with a total energy ∼1061-1062 erg. Using two-dimensional hydrodynamic simulations, we show that if a large fraction of this energy is deposited near the centers of CC clusters, which is likely common due to dense cores, these AGN outbursts can completely remove CCs, transforming them to NCC clusters. Our model also has interesting implications for cluster abundance profiles, which usually show a central peak in CC systems. Our calculations indicate that during the CC to NCC transformation, AGN outbursts efficiently mix metals in cluster central regions and may even remove central abundance peaks if they are not broad enough. For CC clusters with broad central abundance peaks, AGN outbursts decrease peak abundances, but cannot effectively destroy the peaks. Our model may simultaneously explain the contradictory (possibly bimodal) results of abundance profiles in NCC clusters, some of which are nearly flat, while others have strong central peaks similar to those in CC clusters. A statistical analysis of the sizes of central abundance peaks and their redshift evolution may shed interesting insights on the origin of both types of NCC clusters and the evolution history of thermodynamics and AGN activity in clusters.

  4. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  5. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  6. The galaxy cluster RXC J1504-0248: a remarkable cool core near us

    Science.gov (United States)

    Cecília Soja, Ana; Sodre, Laerte; Serra Cypriano, Eduardo; Lima Neto, Gastao B.

    2015-08-01

    One of the most intriguing questions for our understanding of galaxy clusters evolution is the identification of the mechanisms that regulate the temperature of the intergalactic medium in the cluster central region. Conventional cooling models predict a cooling flow of the intracluster gas much higher than observed, suggesting that some mechanisms heats the gas and inhibits the predicted cooling rates. Among the mechanisms proposed for heating the gas, stand out star formation and nuclear activity. The galaxy cluster RXC J1504-0248, located at z = 0.215, is a remarkable example of cool core cluster relatively close to us; its BCG has an extraordinary external filamentary gas structure, comparable to that observed in NGC 1275, in the Perseus cluster. We have studied this object with optical images and spectra obtained with the Gemini South Telescope. We have estimated the cluster mass through weak and strong gravitational lensing techniques, using a NFW-type profile and the analysis of two gravitational arc, respectively. We have obtained 1.5(5) x 10¹⁵ h-1M⊙ within a 3 h-1 Mpc radius (from WL) and 3.74 (3) x 10¹³ M⊙ inside a 62.9 h-1 kpc radius (from SL). These results are consistent with previous estimations obtaneid by Borhringer et al., of 1.7(3) x 10¹⁵ h-1 M⊙ within 3 h-1 Mpc, based in X-ray analysis; the agreement between the WL and X-ray masses estimates is an indication that the cluster is approximately in dynamical equilibrium. Thus, the processes affecting its central region must rely primarily on internal processes to the cluster and, in particular, associated with its dominant galaxy. The analysis of the BCG emission lines with a BPT diagram indicates that, while the nuclear emission is consitent with a LINER, the emission lines from the filamentary region around the BCG comes from star formation.

  7. Abell 262 and RXJ0341: Two Brightest Cluster Galaxies with Line Emission Blanketing a Cool Core

    Science.gov (United States)

    Edwards, Louise O. V.; Heng, Renita

    2014-08-01

    Over the last decade, integral field (IFU) analysis of the brightest cluster galaxies (BCGs) in several cool core clusters has revealed the central regions of these massive old red galaxies to be far from dead. Bright line emission alongside extended X-ray emission links nearby galaxies, is superposed upon vast dust lanes and extends out in long thin filaments from the galaxy core. Yet, to date no unifying picture has come into focus, and the activity across systems is currently seen as a grab-bag of possibile emission line mechanisms. Our primary goal is to work toward a consistent picture for why the BCGs seem are undergoing a renewed level of activity. One problem is most of the current data remains focused on mapping the very core of the BCG, but neglects surrounding galaxies. We propose to discover the full extent of line emission in a complementary pair of BCGs. In Abell 262, an extensive dust patch screens large portions of an otherwise smooth central galaxy, whereas RXJ0341 appears to be a double-core dust free BCG. We will map the full extent of the line emission in order to deduce whether the line emission is a product of local interactions, or the large-scale cluster X-ray gas. The narrow band filter set and large FOV afforded by the the Mayall MOSAIC-1 (MOSA) imager allows us to concurrently conduct an emission line survey of both clusters, locating all line emitting members and beginning a search for the effect of the environment of the different regions (outskirts vs. cluster core) out to the virial radius. We will combine our results with publically available data from 2MASS to determine the upper limits on specific star formation in the BCG and other cluster galaxies within the cluster virial radius.

  8. Lumped parameter analysis of Pb-Bi cooled fast reactor PEACER core using MATRA

    International Nuclear Information System (INIS)

    A lead-bismuth (Pb-Bi) eutectic cooled fast reactor PEACER (Proliferation-resistant, Environment-friendly, Accident-tolerant, Continuable-energy and Economical Reactor) is under development at the Seoul National University. This study is intended to examine the liquid metal coolant behavior along the subchannels and to find out whether the given flux profiles and geometrical arrangement of fuel rods yield reasonable flow distribution during nominal operation using the subchannel analysis code MATRA (Multi-channel Analyzer for Transient and steady-state in Rod Arrays). MATRA was developed at the Korea Atomic Energy Research Institute based on the subchannel approach to calculate the enthalpy and flow distribution in fuel rod bundle elements for both steady-state and transient conditions. The best-estimate analysis was carried out utilizing MATRA for the PEACER-300 quadrant core under the nominal operation condition. Subchannel analysis was performed for the hottest assembly of the PEACER-300 core. The calculation result showed that during normal operation the core material temperature distribution stays well below the thermal design limits. Comparison of the code results with those by hand calculation resulted in good agreement. Hand calculations are in further progress to include the finite difference scheme in the radial direction

  9. Reactivity Effect Of Steam / Water Ingress in Generation-IV Gas-Cooled Fast Reactor Core

    International Nuclear Information System (INIS)

    This paper presents a static neutronic calculational study of steam/water ingress into a Gas-cooled Fast Reactor (GFR Generation IV) core performed by using three Monte-Carlo codes, namely SERPENT version 1.1-16, KENO-VI module of the SCALE, MCNPX version 2.7.0, and different modern nuclear data libraries, i.e. JEFF-3.1, JEFF-3.1.1 and ENDF/B-VII. The analysis was performed for a wide range of water/steam densities [0 – 1.0 g/cm3] within the core and the neutronic parameters were compared between the different codes and libraries. The obtained results demonstrate that this accidental event would result in a large negative reactivity insertion. The main reason of such core behaviour was found to be an increased neutron absorption rate in the cladding liner made of refractory metals (W and Re) due to the neutron spectrum thermalisation resulting from the steam/water ingress. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

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

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

    International Nuclear Information System (INIS)

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

  12. Growth of legionella and other heterotrophic bacteria in a circulating cooling water system exposed to ultraviolet irradiation

    International Nuclear Information System (INIS)

    The effect of ultraviolet irradiation on the growth and occurrence of legionella and other heterotrophic bacteria in a circulating cooling water system was studied. Water of the reservoir was circulated once in 28 h through a side-stream open channel u.v. radiator consisting of two lamps. Viable counts of legionellas and heterotrophic bacteria in water immediately after the u.v. treatment were 0-12 and 0.7-1.2% of those in the reservoir, respectively. U.v. irradiation increased the concentration of easily assimilable organic carbon. In the u.v. irradiated water samples incubated in the laboratory the viable counts of heterotropic bacteria reached the counts in reservoir water within 5 d. The increase in viable counts was mainly due to reactivation of bacterial cells damaged by u.v. light, not because of bacterial multiplication. Despite u.v. irradiation the bacterial numbers in the reservoir water, including legionellas, did not decrease during the experimental period of 33 d. The main growth of bacteria in the reservoir occurred in biofilm and sediment, which were never exposed to u.v. irradiation. (Author)

  13. Scaling Analysis of Natural Circulation Flow Loop

    International Nuclear Information System (INIS)

    To improve the thermal margin for the severe accident measures in high-power reactors, engineered corium cooling systems involving boiling-induced two-phase natural circulation have been proposed for decay heat removal. The boiling-induced natural circulation flow is generated in a coolant path between a hot vessel wall and cold coolant reservoir. In general, an increase in the natural circulation mass flow rate of the coolant leads to an increase in the critical heat flux (CHF) on the hot wall, thus enhancing the thermal margin. An ex-vessel core catcher under consideration, which is one of the engineered corium cooling system, is a passive system consisting of an inclined engineered cooling channel made of a single channel between the body of the core catcher and the inside wall of the reactor cavity. Under severe accident conditions, water is supplied from the IRWST to the engineered cooling channel. The water in the inclined channel absorbs the decay heat transferred from the corium through the carbon steel structure of the core catcher body and boils off as steam. The latter is subsequently released into the free volume of the containment above the corium spreading compartment. Water continues to flow from the IRWST to the cooling channel as a result of buoyancy-driven natural circulation. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. In this study, the scaling analysis was performed by solving the natural circulation flow loop equation for the cooling channel in the ex-vessel core catcher. The scaling analysis was performed by solving the natural circulation flow loop equation for the cooling channel in the ex-vessel core catcher. The boiling-induced natural circulation flow in the cooling channel of the core catcher has been modeled by considering the conservation of mass, momentum and energy in the two-phase mixture, along

  14. Dike control of hydrothermal circulation in the Tertiary Icelandic crust and implications for cooling of the seafloor

    Science.gov (United States)

    Pałgan, Dominik; Devey, Colin W.; Yeo, Isobel A.

    2016-04-01

    Hydrothermal activity along the Mid-Atlantic Ridge is predominantly high-temperature venting controlled by volcano-tectonic processes confined to the ridge axis and neotectonic zone, which extends ~ 20 km on each side of the axis (e.g. TAG or Logatchev 1). These vents cannot, however, account for all the heat which needs to be removed to cool the plate and a significant amount of heat is probably removed in the off-axis regions as well. These regions have previously not been systematically surveyed for hydrothermal activity due to a lack of predictive models for its nature, location or controlling structures. Here we use hot springs in the Tertiary Westfjords of Iceland as onshore analogs for hydrothermal activity along the off-axis Mid-Atlantic Ridge to better understand tectonic and volcanological controls on their occurrence, as well as the processes which support hydrothermal circulation. Our results show that even crust ≥ 10 Ma has abundant low-temperature hydrothermal activity. We show that 66% of hot springs investigated, and 100% of those for which a detailed geological setting could be determined, are associated with basaltic dikes cross-cutting the sub-horizontal lava sequence. This is in strong contrast to on-axis springs, which are known (both from underwater and on land) to be predominantly associated with faults. Absence of earthquakes in Westfjords suggests that the faults there are no longer active and possibly sealed by secondary minerals, suppressing fluid circulation. In such a situation, the jointed and fractures dike margins may provide the major pathways for fluid circulation. Extrapolating this idea to the off-axis regions of the Reykjanes Ridge, we suggest, based on bathymetric maps, potential sites for future exploration for off-axis hydrothermal systems.

  15. An experimental study on two phase natural circulation of external reactor vessel cooling with non-heating method

    International Nuclear Information System (INIS)

    In-Vessel Retention (IVR) concept with External Reactor Vessel Cooling (ERVC) approach has been proved to be effective in removing decay heat from the lower head of Reactor Pressure Vessel (RPV) under severe accident conditions in small- and medium- scale Nuclear Power Plants (NPPs). However, the IVR-ERVC approach still needs to be assessed before its application to large scale NPPs. Heat removal capacity in a large, inverted geometry flow path is highly dependent on the local mass flow rate of natural circulation, which is then affected by various parameters, such as geometry of flow path, height of natural circulation loop, etc. It is desirable to enhance the coolability of ERVC by analyzing and optimizing the parameters affecting mass flow rate and two phase flow behavior. For this purpose, a full-scale, 1-D test facility is designed and set up at Shanghai Jiao Tong University, to study ERVC capability under both natural and forced circulation conditions in large scale NPPs. Two phases of experimental investigation on the facility are projected, emphasis on two-phase flow behavior study of the ERVC flow path applying non-heating method being paid in the first phase, while in the second phase, practical ERVC characteristics investigation being focused using electrically heating approach. This paper reports the first-phase study, in which air injection is used to simulate steam generation, and local and system two-phase flow behaviors, including flowrate trends, bubble transmission and related parametric effects are observed. Test data can be provided for the validation of numerical codes of various classes. (author)

  16. Analysis on non uniform flow in steam generator during steady state natural circulation cooling

    International Nuclear Information System (INIS)

    Steady-state natural circulation (NC) in the PWR was investigated focusing on non uniform flow among steam generator (SG) U-tubes observed in the ROSA/LSTF experiments. In the analysis using the RELAP5/MOD3 code, the SG behavior was analyzed using the partial SG model with one, five, or nine parallel flow paths in the primary side and boundary conditions based on the experiments. The results showed that simulations using the model with five or nine tubes were capable to capture important non uniform phenomena such as reverse flow, fill and dump and stagnant vertical stratification, and the stable SG outlet flow as observed in the experiments. Heat transfer rates to the secondary side were, however, underpredicted by up to 15%. Furthermore, difficulties were found in establishing the steady state condition especially for the low pressure analysis: only when the inlet flow rate was carefully imposed, stable NC behavior was obtained. (author)

  17. CUPID code simulation of the two-phase natural circulation in the passive condensation cooling tank of the PASCAL facility

    International Nuclear Information System (INIS)

    For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. In the present study, the CUPID code was applied for the simulation of the PASCAL (PAFS Condensing Heat Removal Assessment Loop) test facility constructed with an aim of validating the cooling and operational performance of the PAFS (Passive Auxiliary Feedwater System). The PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor +), which is intended to completely replace the conventional active auxiliary feedwater system. This paper introduces the simulation results for the PCCT (passive condensate cooling tank) of the PASCAL facility performed with the CUPID code in order to investigate the thermal hydraulic phenomena in the PCCT. The calculated collapsed water level and local liquid temperature are in good agreement with measured data and the simulation results verified that the important thermal hydraulic characteristics in the PCCT, such as the two-phase natural circulation and the boil-off phenomena, have been successfully reproduced by CUPID. This paper presents the description of the PASCAL test facility, the physical models of the code and its simulation result for the PCCT. (author)

  18. Simulation of single- and two-phase natural circulation in the passive condensate cooling tank using the CUPID code

    International Nuclear Information System (INIS)

    For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal-hydraulic code, named CUPID, has been developed. In the present study, the CUPID code was applied for the simulation of the PASCAL test facility constructed with an aim of validating the cooling and operational performance of the passive auxiliary feedwater system (PAFS). The PAFS is one of the advanced safety features adopted in the Advanced Power Reactor + (APR+), which is intended to completely replace the conventional active auxiliary feedwater system. This paper introduces the simulation results for the passive condensate cooling tank (PCCT) of the PASCAL facility performed with the CUPID code in order to investigate the thermal-hydraulic phenomena in the PCCT. The simulation showed that the important thermal-hydraulic characteristics in the PCCT, such as two-phase natural circulation and boil-off phenomena, can be successfully reproduced by CUPID. Two important validation parameters, collapsed water level and local liquid temperature, were quantitatively well captured in the simulation. This paper presents the description of the PASCAL test facility, the physical models of the CUPID code, and its simulation result for the PCCT. (author)

  19. Core design and safety analyses of 600 MWt, 950 °C high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Masaaki, E-mail: nakano-m@fujielectric.co.jp [Fuji Electric Co., Ltd., 1-1, Tanabe-shinden, Kawasaki-ku, Kawasaki-city 210-9530 (Japan); Takada, Eiji; Tsuji, Nobumasa; Tokuhara, Kazumi; Ohashi, Kazutaka; Okamoto, Futoshi [Fuji Electric Co., Ltd., 1-1, Tanabe-shinden, Kawasaki-ku, Kawasaki-city 210-9530 (Japan); Tazawa, Yujiro; Tachibana, Yukio [Japan Atomic Energy Agency, Oarai, Ibaraki-pref. 311-1393 (Japan)

    2014-05-01

    The conceptual core design study of high temperature gas-cooled reactor (HTGR) is performed. The major specifications are 600 MW thermal output, 950 °C outlet coolant temperature, prismatic core type, enriched uranium fuel. The decay heat in the core can be removed with only passive measures, for example, natural convection reactor cavity cooling system (RCCS), even if any electricity is not supplied (station blackout). The transient thermal analysis of the depressurization accident in the case the primary coolant decreases to the atmosphere pressure shows that the fuels and the reactor pressure vessel temperatures are kept under their safety limit criteria. The fission product release, Ag-110m and Cs-137 from the fuels under the normal operation is small as to make maintenance of devices in the primary cooling system, such as a gas turbine, without remote maintenance. The HTGRs can achieve the advanced safety features based on their inherent passive safety characteristics.

  20. Mathematical Methodology for New Modeling of Water Hammer in Emergency Core Cooling System

    International Nuclear Information System (INIS)

    In engineering insight, the water hammer study has carried out through the experimental work and the fluid mechanics. In this study, a new access methodology is introduced by Newton mechanics and a mathematical method. Also, NRC Generic Letter 2008-01 requires nuclear power plant operators to evaluate the effect of water-hammer for the protection of pipes of the Emergency Core Cooling System, which is related to the Residual Heat Removal System and the Containment Spray System. This paper includes modeling, the processes of derivation of the mathematical equations and the comparison with other experimental work. To analyze the effect of water-hammer, this mathematical methodology is carried out. This study is in good agreement with other experiment results as above. This method is very efficient to explain the water-hammer phenomena

  1. Powering of cool filaments in cluster cores by buoyant bubbles - I. Qualitative model

    Science.gov (United States)

    Churazov, E.; Ruszkowski, M.; Schekochihin, A.

    2013-11-01

    Cool-core clusters (e.g. Perseus or M87) often possess a network of bright gaseous filaments, observed in radio, infrared, optical and X-ray bands. We propose that these filaments are powered by the reconnection of the magnetic field in the wakes of buoyant bubbles. Active galactic nucleus (AGN)-inflated bubbles of relativistic plasma rise buoyantly in the cluster atmosphere, stretching and amplifying the field in the wake to values of β = 8πPgas/B2 ˜ 1. The field lines in the wake have opposite directions and are forced together as the bubble motion stretches the filament. This setup bears strong similarity to the coronal loops on the Sun or to the Earth's magnetotail. The reconnection process naturally explains both the required level of local dissipation rate in filaments and the overall luminosity of filaments. The original source of power for the filaments is the potential energy of buoyant bubbles, inflated by the central AGN.

  2. Powering of cool filaments in cluster cores by buoyant bubbles. I. Qualitative model

    CERN Document Server

    Churazov, E; Schekochihin, A

    2013-01-01

    Cool-core clusters (e.g., Perseus or M87) often possess a network of bright gaseous filaments, observed in radio, IR, optical and X-ray bands. We propose that these filaments are powered by the reconnection of the magnetic field in the wakes of buoyant bubbles. AGN-inflated bubbles of relativistic plasma rise buoyantly in the cluster atmosphere, stretching and amplifying the field in the wake to values of $\\beta =8\\pi P_{gas}/B^2\\sim 1$. The field lines in the wake have opposite directions and are forced together as the bubble motion stretches the filament. This setup bears strong similarity to the coronal loops on the Sun or the Earth magneto-tail. The reconnection process naturally explains both the required level of local dissipation rate in filaments and the overall luminosity of filaments. The original source of power for the filaments is the potential energy of buoyant bubbles, inflated by the central AGN.

  3. Design stability and safety margins of the ETRR-2 core cooling system regarding seismic loads

    International Nuclear Information System (INIS)

    The seismic design of a nuclear power plant includes two levels of design earthquake, the safe shutdown earthquake (SSE) and the operating basis earthquake (OBE). The OBE and SSE are considered in the nuclear power plant design as required by the IAEA safety regulations. A typical piping model for the ETRR-2 core cooling system includes ASME-class 1, 2 and 3 piping and was analyzed with regard to both IAEA safety regimes. A load combination, as stated by the IAEA Safety Guide 50-SG-S2, has been considered. The safety guide requires that the OBE should equal to at least one half SSE and that the plant should be shut down if exposed to earthquake intensity greater than the OBE. The results reflect the requirement for the precise design of a supporting system to accommodate for higher seismic peaks more than 0.2 g and the system needs to be more flexible. (orig.)

  4. Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores

    International Nuclear Information System (INIS)

    In the context of homogenization theory we treat some heat transfer problems involving unusual (according to the homogenization) boundary conditions. These problems are defined in a solid periodic perforated domain where two scales (macroscopic and microscopic) are to be taken into account and describe heat transfer by conduction in the solid and by radiation on the wall of each hole. Two kinds of radiation are considered: radiation in an infinite medium (non-linear problem) and radiation in cavity with grey-diffuse walls (non-linear and non-local problem). The derived homogenized models are conduction problems with an effective conductivity which depend on the considered radiation. Thus we introduce a framework (homogenization and validation) based on mathematical justification using the two-scale convergence method and numerical validation by simulations using the computer code CAST3M. This study, performed for gas cooled reactors cores, can be extended to other perforated domains involving the considered heat transfer phenomena. (author)

  5. Evaluation of instrumentation for detection of inadequate core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    This report is a review of the Approach to Inadequate Core Cooling issue in Boiling Water Reactors (BWR). The report consists of seven sections. The principal conclusion is that the condition of the reference leg, and operator awareness of that condition are of primary importance in level indication reliability for safety. An indication of reference leg level and temperature displayed to the operators would be a useful enhancement of reliability and a guide to further operator action in all circumstances. We conclude that the BWR practice of multiple, redundant coolant level measurements, with overlapping ranges, can be a reliable basis for indication of approach to an ICC condition, and, in correlation with the other control and safety systems of modern BWRs, will prevent unsafe conditions

  6. 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.)

  7. Reproducing cultural identity in negotiating nuclear power: the Union of Concerned Scientists and emergency core cooling

    International Nuclear Information System (INIS)

    This paper advances the concept of 'cultural identity' to account for the nexus between structure and practice in technological negotiations. It describes how the formation of the Union of Concerned Scientists (UCS), and that group's subsequent discourse and nonverbal actions, both reproduced the established identities of group members and contributed to negotiations that reconstituted those identities. In particular, UCS claims about emergency core-cooling systems in nuclear plants were congruent with the combination of a shared ideology, the social interests of Massachusetts Institute of Technology faculty, and established principles of engineering design. The cultural analysis of identity reproduction shows the opposition between cognitive and social phenomena to be a significant distinction framing action in Western culture. The analysis also suggests that new attention be given to the relationship between the constitutive and reproductive functions of discourse and nonverbal action. (author)

  8. The Relation Between Cool Cluster Cores and Herschel-Detected Star Formation in Brightest Cluster Galaxies

    CERN Document Server

    Rawle, T D; Egami, E; Rex, M; Smith, G P; Altieri, B; Fiedler, A; Haines, C P; Pereira, M J; Pérez-González, P G; Portouw, J; Valtchanov, I; Walth, G; van der Werf, P P; Zemcov, M

    2012-01-01

    We present far-infrared (FIR) analysis of 68 Brightest Cluster Galaxies (BCGs) at 0.08 2x10^11 L_sun), only a small (<0.4 mag) reddening correction is required for SFR(Ha) to agree with SFR_FIR. The relatively low Ha extinction (dust obscuration), compared to values reported for the general star-forming population, lends further weight to an alternate (external) origin for the cold gas. Finally, we use a stacking analysis of non-cool-core clusters to show that the majority of the fuel for star formation in the FIR-bright BCGs is unlikely to originate form normal stellar mass loss.

  9. Regulatory assessment of effectiveness of ACR-1000 emergency core cooling system

    International Nuclear Information System (INIS)

    The paper presents the regulatory approach for assessment of the Advanced CANDU Reactor (ACR)-1000 Large Loss of Coolant Accident (LOCA) Emergency Core Cooling (ECC) effectiveness, describes the rationale for the selection of sensitivity cases and discusses the results of the simulations for 50% Pump Suction Break (PSB). The separate in-house simulations strengthened the CNSC staff knowledge about the ACR-1000 design and the modeling methodology. The review of representation of plant systems and plant behavior indicated no major issues. The selected accident scenarios and the limited scope sensitivity cases conducted by the CNSC staff, indicated that, overall, the ECC performance showed small sensitivity to the parameters and assumptions considered for investigation. (author)

  10. Cooling age record of domal uplift in the core of the Higher Himalayan Crystallines (HHC), southwest Zanskar, India

    International Nuclear Information System (INIS)

    The cooling and tectonic history of the Higher Himalayan Crystallines (HHC) in southwest Zanskar (along the Kishtwar-Padam traverse) is constrained by K-Ar biotite and fission-track (FT) apatite and zircon ages. A total of nine biotite samples yields ages in the range of 14-24 Ma, indicating the post-metamorphic cooling of these rocks through ∼ 300 degC in the Miocene. Overall, the ages become younger away from the Zanskar Shear Zone (ZSZ), which marks the basement-cover detachment fault between the HHC and the Tethyan sedimentary zone, towards the core of the HHC. The same pattern is also observed for the FT apatite ages, which record the cooling of the rocks through ∼ 120 degC. The apatite ages range from 11 Ma in the vicinity of the ZSZ to 4 Ma at the granitic core of the HHC. This pattern of discordant cooling ages across the HHC in southwest Zanskar reveals an inversion of isotherms due to fast uplift-denudation (hence cooling) of the HHC core, which is, in turn, related to domal uplift within the HHC. The Chisoti granite gneiss is the exposed domal structure along the studied traverse. Cooling history of two granite gneisses at the core of the HHC is also quantified with the help of the biotite, zircon and apatite ages; the time-temperatures thus obtained indicate a rapid pulse of cooling at ∼ 6 Ma, related to accelerated uplift-denudation of the HHC core at this time. Long-term denudation rates of 0.5-0.7 mm/yr are estimated for the high-grade rocks of the Higher Himalaya in southwest Zanskar over the past 4.0-5.5 m.yr. (author)

  11. Emergency core cooling system sump chemical effects on strainer head loss

    International Nuclear Information System (INIS)

    Chemical precipitates formed in the recovery water following a Loss of Coolant Accident (LOCA) have the potential to increase head loss across the Emergency Core Cooling System (ECCS) strainer, and could lead to cavitation of the ECCS pumps, pump failure and loss of core cooling. AECL, as a strainer vendor and research organization, has been involved in the investigation of chemical effects on head loss for its CANDU® and Pressurized Water Reactor (PWR) customers. The chemical constituents of the recovery sump water depend on the combination of chemistry control additives and the corrosion and dissolution products from metals, concrete, and insulation materials. Some of these dissolution and corrosion products (e.g., aluminum and calcium) may form significant quantities of precipitates. The presence of chemistry control additives such as sodium hydroxide, trisodium phosphate and boric acid can significantly influence the precipitates formed. While a number of compounds may be shown to be thermodynamically possible under the conditions assumed for precipitation, kinetic factors play a large role in the morphology of precipitates. Precipitation is also influenced by insulation debris, which can trap precipitates and act as nucleation sites for heterogeneous precipitation. This paper outlines the AECL approach to resolving the issue of chemical effects on ECCS strainer head loss, which included modeling, bench top testing and reduced-scale testing; the latter conducted using a temperature-controlled variable-flow closed-loop test rig that included an AECL Finned Strainer® test section equipped with a differential pressure transmitter. Models of corrosion product release and the effects of precipitates on head loss will also be presented. Finally, this paper discusses the precipitates found in test debris beds and presents a possible method for chemical effects head loss modeling. (author)

  12. New Detections of Radio Minihalos in Cool Cores of Galaxy Clusters

    Science.gov (United States)

    Giacintucci, Simona; Markevitch, Maxim; Venturi, Tiziana; Clarke, Tracy E.; Cassano, Rossella; Mazzotta, Pasquale

    2013-01-01

    Cool cores of some galaxy clusters exhibit faint radio minihalos. Their origin is unclear, and their study has been limited by their small number. We undertook a systematic search for minihalos in a large sample of X-ray luminous clusters with high-quality radio data. In this article, we report four new minihalos (A 478, ZwCl 3146,RXJ 1532.9+3021, and A 2204) and five candidates found in the reanalyzed archival Very Large Array observations.The radio luminosities of our minihalos and candidates are in the range of 102325 W Hz1 at 1.4 GHz, which is consistent with these types of radio sources. Their sizes (40160 kpc in radius) are somewhat smaller than those of previously known minihalos. We combine our new detections with previously known minihalos, obtaining a total sample of 21 objects, and briefly compare the cluster radio properties to the average X-ray temperature and the total masses estimated from Planck.We find that nearly all clusters hosting minihalos are hot and massive. Beyond that, there is no clear correlation between the minihalo radio power and cluster temperature or mass (in contrast with the giant radio halos found in cluster mergers, whose radio luminosity correlates with the cluster mass). Chandra X-ray images indicate gas sloshing in the cool cores of most of our clusters, with minihalos contained within the sloshing regions in many of them. This supports the hypothesis that radio-emitting electrons are reaccelerated by sloshing. Advection of relativistic electrons by the sloshing gas may also play a role in the formation of the less extended minihalos.

  13. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  14. Kinetic AGN Feedback Effects on Cluster Cool Cores Simulated using SPH

    CERN Document Server

    Barai, Paramita; Borgani, Stefano; Gaspari, Massimo; Granato, Gian Luigi; Monaco, Pierluigi; Ragone-Figueroa, Cinthia

    2016-01-01

    We implement novel numerical models of AGN feedback in the SPH code GADGET-3, where the energy from a supermassive black hole (BH) is coupled to the surrounding gas in the kinetic form. Gas particles lying inside a bi-conical volume around the BH are imparted a one-time velocity (10,000 km/s) increment. We perform hydrodynamical simulations of isolated cluster (total mass 10^14 /h M_sun), which is initially evolved to form a dense cool core, having central T<10^6 K. A BH resides at the cluster center, and ejects energy. The feedback-driven fast wind undergoes shock with the slower-moving gas, which causes the imparted kinetic energy to be thermalized. Bipolar bubble-like outflows form propagating radially outward to a distance of a few 100 kpc. The radial profiles of median gas properties are influenced by BH feedback in the inner regions (r<20-50 kpc). BH kinetic feedback, with a large value of the feedback efficiency, depletes the inner cool gas and reduces the hot gas content, such that the initial c...

  15. Chandra Observation of Abell 1142: A Cool-core Cluster Lacking a Central Brightest Cluster Galaxy?

    Science.gov (United States)

    Su, Yuanyuan; Buote, David A.; Gastaldello, Fabio; van Weeren, Reinout

    2016-04-01

    Abell 1142 is a low-mass galaxy cluster at low redshift containing two comparable brightest cluster galaxies (BCGs) resembling a scaled-down version of the Coma Cluster. Our Chandra analysis reveals an X-ray emission peak, roughly 100 kpc away from either BCG, which we identify as the cluster center. The emission center manifests itself as a second beta-model surface brightness component distinct from that of the cluster on larger scales. The center is also substantially cooler and more metal-rich than the surrounding intracluster medium (ICM), which makes Abell 1142 appear to be a cool-core cluster. The redshift distribution of its member galaxies indicates that Abell 1142 may contain two subclusters, each of which contain one BCG. The BCGs are merging at a relative velocity of ≈1200 km s-1. This ongoing merger may have shock-heated the ICM from ≈2 keV to above 3 keV, which would explain the anomalous LX-TX scaling relation for this system. This merger may have displaced the metal-enriched “cool core” of either of the subclusters from the BCG. The southern BCG consists of three individual galaxies residing within a radius of 5 kpc in projection. These galaxies should rapidly sink into the subcluster center due to the dynamical friction of a cuspy cold dark matter halo.

  16. Rhapsody-G simulations I: the cool cores, hot gas and stellar content of massive galaxy clusters

    CERN Document Server

    Hahn, Oliver; Wu, Hao-Yi; Evrard, August E; Teyssier, Romain; Wechsler, Risa H

    2015-01-01

    We present the Rhapsody-G suite of cosmological hydrodynamic AMR zoom simulations of ten massive galaxy clusters at the $M_{\\rm vir}\\sim10^{15}\\,{\\rm M}_\\odot$ scale. These simulations include cooling and sub-resolution models for star formation and stellar and supermassive black hole feedback. The sample is selected to capture the whole gamut of assembly histories that produce clusters of similar final mass. We present an overview of the successes and shortcomings of such simulations in reproducing both the stellar properties of galaxies as well as properties of the hot plasma in clusters. In our simulations, a long-lived cool-core/non-cool core dichotomy arises naturally, and the emergence of non-cool cores is related to low angular momentum major mergers. Nevertheless, the cool-core clusters exhibit a low central entropy compared to observations, which cannot be alleviated by thermal AGN feedback. For cluster scaling relations we find that the simulations match well the $M_{500}-Y_{500}$ scaling of Planck ...

  17. A numerical model of the effects of reactor cooling water on fjord circulation. Part 2, figures

    International Nuclear Information System (INIS)

    In the search for possible sites for new nuclear power plants in Sweden a site on Braaviken, a narrow fjord, is being considered. A numerical hydrodynamic model has been developed to predict the probable effects of the waste heat disharged into the estuary on the natural estuarine flow. The model employs the basic equations of motion and conservation of salt and heat with appropriate approximations to make predictions. The primary approximation in the model consists of considering the estuary as a channel in which cross channel effects do not explicitly appear. The along channel motion is thus primary determined by the along channel density gradients. With the construction of a bottom intake located at the depth of about 40 meters there will be little noticable effect on the circulation, temperature or salinity fields in the estuary in the summer. However in the winter the bottom intake offers only a partial improvement over a surface intake. During the winter the heated water would cause changes of as much as 50 % in the natural state. The surface intake would cause changes which sometimes are almost twice as big. The problem arises because the 10 deg C heated water creates sizable horizontal density gradients which are sufficient to counteract the weak natural flow

  18. Cooling system of the core of a nuclear reactor while it is being stopped or normally operating

    International Nuclear Information System (INIS)

    The present invention proposes a cooling system with intermediate gas flow which ensures the reactor core cooling when the primary pumps are stopped either directly by means of main heat-exchange circuits ensuring normally the reactor operation, or by means of separated loops, these ones being able so to operate in an autonomous way for they produce their own electricity needs and also an excedent which is added to the power plant production. The cooling circuit and the heat exchanger are described in detail

  19. VARIAN加速器内循环水冷维修%Maintenance of Internal Circulation Water Cooling System of VARIAN Accelerator

    Institute of Scientific and Technical Information of China (English)

    逄宏义

    2012-01-01

    介绍了VARIAN(瓦里安)加速器内循环水冷系统的原理及故障维修。%Introduce the working principles and trouble shooting of the internal circulation water cooling system of VARIAN Accelerator.

  20. A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    Highlights: • SFR burner core configurations are explored and analyzed for effective use of thorium blankets. • Thorium blankets can significantly improve SFR burner core performances. • No recycling or partial recycling of Th blankets with multi-batches is very effective. - Abstract: In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate

  1. Trends in development of in-core monitoring of light water cooled reactor power distribution

    International Nuclear Information System (INIS)

    Trends and achievements in the development of in-core monitoring of light water cooled reactor power distribution are described. Application of inertialess detectors in control systems of the 1300 MW PWR reactors operating in the load-following regime, various options of the systems, the capabilities and prospects of further improvement are considered. Problems accompanying their application are analysed. Wire and fall-type activation detectors were used in the PWR and BWR reactors at the initial stages of in-core monitoring system development. Later systems with stationary β-emission neutron detectors, with rhodium or vanadium emitters and compton emission neutron detectors with emitters of cobalt or hafnium gained wide application. Gamma thermometer which require no calibration during the operation appear to be new detectors for continuous control in PWRs. However they are intertice and their application does not rid of the necessity to have sensors with minor signal delay in the control system and does not allow one to refuse periodic changes of axial power by mobile detectors, providing for high accuracy and detalization of distributions

  2. Steam generation: fossil-fired systems: utility boilers; industrial boilers; boiler auxillaries; nuclear systems: boiling water; pressurized water; in-core fuel management; steam-cycle systems: condensate/feedwater; circulating water; water treatment

    International Nuclear Information System (INIS)

    A survey of development in steam generation is presented. First, fossil-fired systems are described. Progress in the design of utility and industrial boilers as well as in boiler auxiliaries is traced. Improvements in coal pulverizers, burners that cut pollution and improve efficiency, fans, air heaters and economisers are noted. Nuclear systems are then described, including the BWR and PWR reactors, in-core fuel management techniques are described. Finally, steam-cycle systems for fossil-fired and nuclear power plants are reviewed. Condensate/feedwater systems, circulating water systems, cooling towers, and water treatment systems are discussed

  3. Preliminary design study of a board type radial fuel shuffling sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Highlights: • A 1500MWt radial fuel shuffling sodium cooled breed and burn reactor core was designed. • The board type radial fuel shuffling strategy was applied and demonstrated. • Influences of the fuel height and core radius were investigated. - Abstract: In this paper, a preliminary board type radial fuel shuffling sodium cooled breed and burn reactor core is designed. In the current design, a number of breeding subassemblies are arranged in the center core to ensure enough breeding. A self-developed MCNP-ORIGEN coupled system with the ENDF/B-VI data library is applied to perform neutronics and burn-up calculations. For a 2.0 m radius and 2.5 m height core, the results demonstrate the feasibility of the board type radial fuel shuffling strategy. Breeding mainly occurs in the breeding subassemblies during the first 6 fuel cycles as they are moved to the burning/breeding region. The core will become asymptotically stable after about 24 years. The discharged burn-up of most subassemblies is about 15.0–30.0%. The influences of the core size on the major core parameters, such as initial keff, steady keff, maximum power density, peak burn-up and burn-up ratio between breeding and ignition subassemblies are calculated and investigated. The results indicate that the initial keff increases with fuel height and core radius and finally reaches stability; the steady keff increases with fuel height and core radius, then reaches peak value and finally decreases; the maximum power density, the peak burn-up and the burn-up ratio between breeding and ignition subassemblies decrease with the increase of fuel height and core radius; if core radius is less than 1.875 m, they increase sharply with the decrease of core radius

  4. A 290-a record of atmospheric circulation over the North Pacific from a Mt. Logan ice core, Yukon Territory

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    Calibrations between sodium (Na+) concentrations from a Mt. Logan ice core and sea level pressure (SLP) series show that Na+ concentrations are closely correlated with the autumn-time (SeptemberOctober-November) Aleutian low (AleuLow). A deepening of the AleuLow strengthens the transport of sea-salt aerosols from the North Pacific to the Mt. Logan region. The Mt. Logan Na+ record is used to develop a 292 a (1688~1979) reconstruction of the AleuLow revealing a dramatic intensification of atmospheric circulation over the North Pacific region since the 20th century. Mean SLP of the AleuLow was about 1 hPa lower during the 20th century than during prior periods. The strongest deepening of the AleuLow appeared in the 1950s. Significant correlations are also found between the Mt. Logan AleuLow proxy series and the Pacific decadal oscillation (PDO) and Pacific circulation (PC) index during the 20th century. Evolutionary spectral analysis of the proxy record shows significant periodicities from 15 to 30 a consistent with PDO fluctuations and the bidecadal oscillation of North Pacific atmosphere-ocean circulation. A period of 11 a in the AleuLow record may be associated with the Schwabe 11-a cycle of sunspot activity. Additional longer ice core records from this region will aid in the efforts to further understand the climatic change over the North Pacific region.

  5. Core temperatures during major abdominal surgery in patients warmed with new circulating-water garment, forced-air warming, or carbon-fiber resistive-heating system

    OpenAIRE

    Hasegawa, Kenji; Negishi, Chiharu; Nakagawa, Fumitoshi; Ozaki, Makoto

    2011-01-01

    Purpose It has been reported that recently developed circulating-water garments transfer more heat than a forced-air warming system. The authors evaluated the hypothesis that circulating-water leg wraps combined with a water mattress better maintain intraoperative core temperature ≥36°C than either forced-air warming or carbon-fiber resistive heating during major abdominal surgery. Methods Thirty-six patients undergoing open abdominal surgery were randomly assigned to warming with: (1) circul...

  6. European Lead-cooled SYstem core design: an approach towards sustainability

    International Nuclear Information System (INIS)

    This paper deals with the neutronic design of ELSY (the European Lead-cooled SYstem), a 600 MWe Fast Reactor developed within the 6th EURATOM Framework Programme. The overall core layout, characterized by open square Fuel Assemblies in a rectangular staggered lattice configuration mostly defined by complying with mechanical and seismic constraints, has been optimized in order to obtain a flat power/Fuel Assembly distribution (maximum-to-average ratio: 1.2). The power-to-flow ratio is locally adjusted by changing the fissile (Plutonium) enrichment at different radial positions in the core. Three independent scram systems have been introduced in order to achieve the required reliability for reactor shutdown and safety: eight traditional concept Control Rod assemblies together with two sets of sparse control 'Finger' Absorber Rods, small B4C rods that can be inserted, in principle, in the centre of each FA. One of the two finger absorber systems includes a motorized subset devoted to the regulation of the criticality swing during the cycle: their number can be limited indeed since the small reactivity swing (some hundreds pcm) due to the about unitary breeding ratio. Such an innovative solution can also be positioned in order to maintain an optimal power flattening during the fuel cycle. The core design of ELSY has been organized aiming also at showing that it is possible to realize an 'adiabatic' reactor, i.e. a reactor self-sustainable in Plutonium and burning its own generated Minor Actinides. This complies with the sustainability goal of Generation IV systems: for the implementation of a closed fuel cycle the forthcoming reactors would have to base their operation upon the net 'conversion' of either Natural or Depleted Uranium into Fission Products only. (author)

  7. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  8. Analysis of natural circulation in the in-core structure test section (T2) in the case of a blower trip

    International Nuclear Information System (INIS)

    When a blower trip occurs in an abnormal condition of the in-core structure test section (T2), natural circulation will develop in the two flow channels which are formed by the gap between the fixed reflector and the side shield and the gap between the side shield and the core barrel. The natural circulation heats up the structures of T2, such as a core restraint mechanism and a core barrel and others. Moreover, the radiation emitted from the heated core barrel enhances markably heating-up of the pressure vessel. This report deals with an analysis of the natural circulation accurred after a blower trip, and with the effect on the temperature rise of the structures of T2. Possible countermeasures are also discussed. (author)

  9. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  10. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki [Nuclear Research Group, FMIPA, Bandung Institute of Technology Jl. Ganesha 10, Bandung 40132 (Indonesia); Miura, Ryosuke; Takaki, Naoyuki [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, H. [Emerritus Prof. of Research Laboratory for Nuclear Reactors, Tokyo Inst. of Technology (Japan)

    2015-09-30

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  11. A computer code for analysis of core transient behavior in a Na-cooled metal fuel fast reactor

    International Nuclear Information System (INIS)

    The core transient behavior calculation code 'EXCURS' for a Na-cooled oxide fuel fast reactor was modified for the application to a Na-cooled metal fuel fast reactor (LMR). The results of the core transient behavior calculated with the modified EXCURS were compared with those calculated by ANL for EBR-II and also compared with those by CRIEPI for 1000MWe-LMR. These calculations agreed quite well. The modified EXCURS, therefore, can be used for analysing the core transient behavior of LMR. In a design study of actinide burner reactors (ABR), the analysis of core transient behavior is important from the viewpoint of safety. The ULOF and UTOP analyses for a Na-cooled metal fuel ABR (M-ABR) were carried out using the modified EXCURS. The effect of heat conductivity of fuel and that of feedback reactivity coefficients on the core transient behavior were also evaluated. It is calculated that the maximum temperature of fuel is strongly affected by flowering reactivity coefficient, delayed neutron fraction and heat conductivity of fuel in this order. (author)

  12. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    International Nuclear Information System (INIS)

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis

  13. Pre-conceptual study of small modular PbBi-cooled nitride fuel reactor core characteristics

    International Nuclear Information System (INIS)

    Highlights: • The nitride fuel, stainless steel cladding and Pb-Bi coolant have a perfect compatibility with each other, as well as with excellent neutronics characteristics. • High conversion ratios have been achieved by optimizing the designed parameters to ensure 20 EFPY without refueling and shuffling. • The burn-up swings slightly due to the perfect breeding capability. • All of the important reactivity coefficients are negative to assure SMoPN holding passive safety. • The control system can provide the enough shutdown margins in any operational conditions. - Abstract: In this paper a pre-conceptual neutronics study on a small modular Pb-Bi cooled reactor with nitride fuel (SMoPN) is presented. The SMoPN is designed to meet the requirements for nuclear energy expansion in the next decades, by using the plutonium and thorium nitride fuel to increase the efficiency and performance of fuel. Based on the existing experiences of nuclear reactor, the primary design parameters were provided to match the design goals by the whole core three-dimensional calculation. The nitride fuel, stainless steel cladding and Pb-Bi coolant have a perfect compatibility with each other, as well as with excellent neutronics characteristics. High conversion ratios have been achieved to ensure 20 effective full power years (EFPYs) without refueling and shuffling. During the core lifetime, the burn-up swings slightly due to the perfect breeding capability. All of the important reactivity coefficients are negative to assure the SMoPN holding passive safety. The control system can provide enough shutdown margins in both normal and abnormal operational conditions. Therefore, the SMoPN concept satisfies completely the advanced design idea and the requirements of advanced nuclear reactor system

  14. Experimental investigation of natural circulation BWR core-wide and regional stability on the basis of time series analysis

    International Nuclear Information System (INIS)

    A time series analysis method was performed to calculate decay ratios from dominant poles of a transfer function by applying AR method to time series of the core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excessively conservative conditions. Experiments were conducted with the SIRIUS facility, which simulates a representative natural circulation BWR. Channel and regional stability decay ratios at the nominal operating condition were determined to be 0.38 and 0.54, respectively, which indicates sufficient margin for the instabilities. Experiments were extended to investigate the effects of the design parameters on stability. For the marginal operating condition, the system further stabilized with decreasing a ratio of outer to inner power, a core inlet subcooling, and void reactivity coefficient. The system became the least stable condition when thermal conductance of the fuel rod coincided with the oscillation period of thermal-hydraulic instability. (author)

  15. MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

    International Nuclear Information System (INIS)

    The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR

  16. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    International Nuclear Information System (INIS)

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company's Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates

  17. Application of Full Equation in Kinematic Shock on Emergency Core Cooling System

    International Nuclear Information System (INIS)

    Generic Letter 2008-01shows that the void packet inner of pipes in front of ECCS(Emergency Core Cooling System) pumps is very important effect element in analyzing head loss. The purpose of this paper is to develop the solution of the kinematic shock equation. In this work, the simplified equation of Perdu test study is changed into Full equation by our development study of calculating a kinematic shock. The development result of the solution of full equation is applied and compared with current simplified equation. Finally, the full equation method is used for calculating the criteria of void packet in Westinghouse type NPP in preliminary sample study. In this work's theoretical base is on the study of Seung-Chan Lee et al in KHNP in 2012. The new method of calculating the depth of the kinematic shock in U-type pipe in ECCS is introduced. The kinematic shock is strongly depended on the void packet velocity. In the part of verification, the difference between this work and Perdu experiment result is nothing in the condition of many iterations of Perdu simple model. In conclusion, this work's method is more efficient than Perdu simple model because of the use of only one-step calculation. In void packet criteria, using full equation, some results are calculated. The results are ranged from 0.3 ft''3 to 6.12ft''3 in Westinghouse type NPP

  18. New XMM-Newton observation of the Phoenix cluster: properties of the cool core

    CERN Document Server

    Tozzi, P; Molendi, S; Ettori, S; Santos, J S; De Grandi, S; Balestra, I; Rosati, P; Altieri, B; Cresci, G; Menanteau, F; Valtchanov, I

    2015-01-01

    (Abridged) We present a spectral analysis of a deep (220 ks) XMM-Newton observation of the Phoenix cluster (SPT-CL J2344-4243), which we also combine with Chandra archival ACIS-I data. We extract CCD and RGS X-ray spectra from the core region to search for the signature of cold gas, and constrain the mass deposition rate in the cooling flow which is thought to be responsible of the massive star formation episode observed in the BCG. We find an average mass deposition rate of $\\dot M = 620 (-190 +200)_{stat} (-50 +150)_{syst} M_\\odot$/yr in the temperature range 0.3-3.0 keV from MOS data. A temperature-resolved analysis shows that a significant amount of gas is deposited only above 1.8 keV, while upper limits of the order of hundreds of $M_\\odot$/yr can be put in the 0.3-1.8 keV temperature range. From pn data we obtain $\\dot M = 210 (-80 +85)_{stat} ( -35 +60)_{syst} M_\\odot$/yr, and the upper limits from the temperature-resolved analysis are typically a factor of 3 lower than MOS data. In the RGS spectrum, n...

  19. Application of Full Equation in Kinematic Shock on Emergency Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung-Chan; Yoon, Duk-Joo; Ha, Sang-Jun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Generic Letter 2008-01shows that the void packet inner of pipes in front of ECCS(Emergency Core Cooling System) pumps is very important effect element in analyzing head loss. The purpose of this paper is to develop the solution of the kinematic shock equation. In this work, the simplified equation of Perdu test study is changed into Full equation by our development study of calculating a kinematic shock. The development result of the solution of full equation is applied and compared with current simplified equation. Finally, the full equation method is used for calculating the criteria of void packet in Westinghouse type NPP in preliminary sample study. In this work's theoretical base is on the study of Seung-Chan Lee et al in KHNP in 2012. The new method of calculating the depth of the kinematic shock in U-type pipe in ECCS is introduced. The kinematic shock is strongly depended on the void packet velocity. In the part of verification, the difference between this work and Perdu experiment result is nothing in the condition of many iterations of Perdu simple model. In conclusion, this work's method is more efficient than Perdu simple model because of the use of only one-step calculation. In void packet criteria, using full equation, some results are calculated. The results are ranged from 0.3 ft''3 to 6.12ft''3 in Westinghouse type NPP.

  20. Core flows and heat transfer induced by inhomogeneous cooling with sub- and supercritical convection

    CERN Document Server

    Dietrich, Wieland; Wicht, Johannes

    2016-01-01

    The amount and spatial pattern of heat extracted from cores of terrestrial planets is ultimately controlled by the thermal structure of the lower rocky mantle. Using the most common model to tackle this problem, a rapidly rotating and differentially cooled spherical shell containing an incompressible and viscous liquid is numerically investigated. To gain the physical basics, we consider a simple, equatorial symmetric perturbation of the CMB heat flux shaped as a spherical harmonic $Y_{11}$. The thermodynamic properties of the induced flows mainly depend on the degree of nonlinearity parametrised by a horizontal Rayleigh number $Ra_h=q^\\ast Ra$, where $q^\\ast$ is the relative CMB heat flux anomaly amplitude and $Ra$ is the Rayleigh number which controls radial buoyancy-driven convection. Depending on $Ra_h$ we characterise three flow regimes through their spatial patterns, heat transport and flow speed scalings: in the linear conductive regime the radial inward flow is found to be phase shifted $90^\\circ$ eas...

  1. X-Ray cavities and temperature jumps in strong cool core cluster Abell 2390

    CERN Document Server

    Sonkamble, S S; Pawar, P K; Patil, M K

    2014-01-01

    We present results based on the systematic analysis of high resolution 95\\,ks \\textit{Chandra} observations of the strong cool core cluster Abell 2390 at the redshift of z = 0.228, which hosts an energetic radio AGN. This analysis has enabled us to investigate five X-ray deficient cavities in the hot atmosphere of Abell 2390 within central 30\\arcsec, three of which are newly detected. Presence of these cavities have been confirmed through a various image processing techniques like, the surface brightness profiles, unsharp masked image, as well as 2D elliptical model subtracted residual map. Temperature profile as well as 2D temperature map revealed structures in the distribution of ICM, in the sense that ICM in NW direction is relatively cooler than that on the SE direction. Two temperature jumps, one from 6\\,keV to 9.25\\,keV at 72 kpc on the north direction, and the other from 6\\,keV to 10.27\\,keV at 108 kpc in the east direction have been observed. These temperature jumps are associated with the shocks with...

  2. Central Mass Profiles of the Nearby Cool-core Galaxy Clusters Hydra A and A478

    CERN Document Server

    Okabe, N; Tamura, T; Fujita, Y; Takizawa, M; Matsushita, K; Fukazawa, Y; Futamase, T; Kawaharada, M; Miyazaki, S; Mochizuki, Y; Nakazawa, K; Ohashi, T; Ota, N; Sasaki, T; Sato, K; Tam, S I

    2015-01-01

    We perform a weak-lensing study of the nearby cool-core galaxy clusters, Hydra A ($z=0.0538$) and A478 ($z=0.0881$), of which brightest cluster galaxies (BCGs) host powerful activities of active galactic nuclei (AGNs). For each cluster, the observed tangential shear profile is well described either by a single Navarro--Frenk--White model or a two-component model including the BCG as an unresolved point mass. For A478, we determine the BCG and its host-halo masses from a joint fit to weak-lensing and stellar photometry measurements. We find that the choice of initial mass functions (IMFs) can introduce a factor of two uncertainty in the BCG mass, whereas the BCG host halo mass is well constrained by data. We perform a joint analysis of weak-lensing and stellar kinematics data available for the Hydra A cluster, which allows us to constrain the central mass profile without assuming specific IMFs.We find that the central mass profile ($r<300$ kpc) determined from the joint analysis is in excellent agreement wi...

  3. Biased total mass of cool core galaxy clusters by Sunyaev-Zel'dovich Effect measurements

    CERN Document Server

    Conte, A; Comis, B; Lamagna, L; De Gregori, S

    2010-01-01

    The Sunyaev Zel'dovich (SZ) effect is one of the most powerful cosmological tools to investigate the large-scale Universe, in which clusters of galaxies are the most interesting target. The great advantage of the SZ effect of being redshift independent, in contrast with visible and X-ray observations, allows to directly estimate cluster total mass from the integrated comptonization parameter Y, even for faraway clusters. However, the lack of a complete knowledge of the Intra-Cluster gas (ICg) physics can affect the results. Taking into account self-similar temperature and density profiles of the ICg, we study how different ICg morphologies can affect the cluster total mass estimation. Due to the large percentage of cool core (CC) clusters, we analyze this class starting with a limited sample of eight objects, observed by Chandra. We simulate SZ observations of these clusters through X-ray derived information, and re-analyze the mock SZ data with the simplistic assumption for the ICg of an isothermal beta mode...

  4. Application of natural circulation systems: advantages and challenges - II

    International Nuclear Information System (INIS)

    Applications of natural circulation systems are provided for advanced light water reactor designs. Design features proposed for the passive advanced light water reactors include the use of passive, gravity-fed water supplies for emergency core cooling and natural circulation decay heat removal from the primary system and the containment, and natural circulation cooling within the core for all conditions. Examples are given from different types of advanced reactor designs for the use of passive safety systems under the operational, transient, and accident conditions. Challenges encountered in the design of passive safety systems for HPLWR are discussed in short, as an example case. (author)

  5. New XMM-Newton observation of the Phoenix cluster: properties of the cool core

    Science.gov (United States)

    Tozzi, P.; Gastaldello, F.; Molendi, S.; Ettori, S.; Santos, J. S.; De Grandi, S.; Balestra, I.; Rosati, P.; Altieri, B.; Cresci, G.; Menanteau, F.; Valtchanov, I.

    2015-08-01

    Aims: We present a spectral analysis of a deep (220 ks) XMM-Newton observation of the Phoenix cluster (SPT-CL J2344-4243). We also use Chandra archival ACIS-I data that are useful for modeling the properties of the central bright active galactic nucleus and global intracluster medium. Methods: We extracted CCD and reflection grating spectrometer (RGS) X-ray spectra from the core region to search for the signature of cold gas and to finally constrain the mass deposition rate in the cooling flow that is thought to be responsible for the massive star formation episode observed in the brightest cluster galaxy (BCG). Results: We find an average mass-deposition rate of Ṁ = 620 (-190 + 200)stat (-50 + 150)syst M⊙ yr-1 in the temperature range 0.3-3.0 keV from MOS data. A temperature-resolved analysis shows that a significant amount of gas is deposited at about 1.8 keV and above, while only upper limits on the order of hundreds of M⊙ yr-1 can be placed in the 0.3-1.8 keV temperature range. From pn data we obtain Ṁ = 210 (-80 + 85)stat (-35 + 60)syst M⊙ yr-1 in the 0.3-3.0 keV temperature range, while the upper limits from the temperature-resolved analysis are typically a factor of 3 lower than MOS data. No line emission from ionization states below Fe XXIII is seen above 12 Å in the RGS spectrum, and the amount of gas cooling below ~3 keV has a formal best-fit value Ṁ = 122-122+343 M⊙ yr-1. In addition, our analysis of the far-infrared spectral energy distribution of the BCG based on Herschel data provides a star formation rate (SFR) equal to 530 M⊙ yr-1 with an uncertainty of 10%, which is lower than previous estimates by a factor 1.5. Overall, current limits on the mass deposition rate from MOS data are consistent with the SFR observed in the BCG, while pn data prefer a lower value of Ṁ ~ SFR/ 3, which is inconsistent with the SFR at the 3σ confidence level. Conclusions: Current data are able to firmly identify a substantial amount of cooling gas only

  6. Association between atmospheric circulation patterns and firn-ice core records from the Inilchek glacierized area, central Tien Shan, Asia

    Science.gov (United States)

    Aizen, V.B.; Aizen, E.M.; Melack, J.M.; Kreutz, K.J.; Cecil, L.D.

    2004-01-01

    Glacioclimatological research in the central Tien Shan was performed in the summers of 1998 and 1999 on the South Inilchek Glacier at 5100-5460 m. A 14.36 m firn-ice core and snow samples were collected and used for stratigraphic, isotopic, and chemical analyses. The firn-ice core and snow records were related to snow pit measurements at an event scale and to meteorological data and synoptic indices of atmospheric circulation at annual and seasonal scales. Linear relationships between the seasonal air temperature and seasonal isotopic composition in accumulated precipitation were established. Changes in the ??18O air temperature relationship, in major ion concentration and in the ratios between chemical species, were used to identify different sources of moisture and investigate changes in atmospheric circulation patterns. Precipitation over the central Tien Shan is characterized by the lowest ionic content among the Tien Shan glaciers and indicates its mainly marine origin. In seasons of minimum precipitation, autumn and winter, water vapor was derived from the and and semiarid regions in central Eurasia and contributed annual maximal solute content to snow accumulation in Tien Shan. The lowest content of major ions was observed in spring and summer layers, which represent maximum seasonal accumulation when moisture originates over the Atlantic Ocean and Mediterranean and Black Seas. Copyright 2004 by the American Geophysical Union.

  7. A multi-wavelength view of cooling vs. AGN heating in the X-ray luminous cool-core of Abell 3581

    CERN Document Server

    Canning, R E A; Sanders, J S; Clarke, T E; Fabian, A C; Giacintucci, S; Lal, D V; Werner, N; Allen, S W; Donahue, M; Johnstone, R M; Nulsen, P E J; Sarazin, C L

    2013-01-01

    We report the results of a multi-wavelength study of the nearby galaxy group, Abell 3581 (z=0.0218). This system hosts the most luminous cool core of any nearby group and exhibits active radio mode feedback from the super-massive black hole in its brightest group galaxy, IC 4374. The brightest galaxy has suffered multiple active galactic nuclei outbursts, blowing bubbles into the surrounding hot gas, which have resulted in the uplift of cool and cold gas into the surrounding hot intragroup medium. High velocities, indicative of an outflow, are observed close to the nucleus and coincident with the radio jet. Thin dusty filaments accompany the uplifted, ionised gas. No extended star formation is observed, however, a young cluster is detected just north of the nucleus. The direction of rise of the bubbles has changed between outbursts. This directional change is likely due to sloshing motions of the intragroup medium. These sloshing motions also appear to be actively stripping the X-ray cool core, as indicated b...

  8. Conceptual design study of an accelerator-based actinide transmutation plant with sodium-cooled solid target/core

    International Nuclear Information System (INIS)

    Research and development works on accelerator-based nuclear waste transmutation are carried out at JAERI under the national program OMEGA. The preliminary design of the proposed minor actinide transmutation plant with a solid target/core is described. The plant consists of a high intensity proton accelerator, spallation target of solid tungsten, and subcritical core loaded with actinide alloy fuel. Minor actinides are transmuted by fast fission reactions. The target and core are cooled by the forced flow of liquid sodium coolant. Thermal energy is recovered to supply electricity to power its own accelerator. The core with an effective multiplication factor of about 0.9 generates. The thermal power of 820 MW by using a 1.5 GeV proton beam with a current of 39 mA. The average burnup is about 8%, about 250 kg of actinides, after one year operation at an 80% of load factor. With the conventional steam turbine cycle, electric output of about 246 MW is produced. The design of the transmutation plant with sodium-cooled solid target/core is mostly based on the well-established technology of current LMFRs. Advantages and disadvantages of solid target/core are discussed. Recent progress in the development of intense proton accelerator, the development of simulation code system, and the spallation integral experiment is also presented. (author)

  9. Casting core for a cooling arrangement for a gas turbine component

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ching-Pang; Heneveld, Benjamin E

    2015-01-20

    A ceramic casting core, including: a plurality of rows (162, 166, 168) of gaps (164), each gap (164) defining an airfoil shape; interstitial core material (172) that defines and separates adjacent gaps (164) in each row (162, 166, 168); and connecting core material (178) that connects adjacent rows (170, 174, 176) of interstitial core material (172). Ends of interstitial core material (172) in one row (170, 174, 176) align with ends of interstitial core material (172) in an adjacent row (170, 174, 176) to form a plurality of continuous and serpentine shaped structures each including interstitial core material (172) from at least two adjacent rows (170, 174, 176) and connecting core material (178).

  10. Core support performance test in the component flow test loop

    International Nuclear Information System (INIS)

    A description is given of the Core Flow Test Loop at Oak Ridge National Laboratory. This is a closed-circuit, out-of-pile loop circulating helium at temperatures and pressures anticipated in gas-cooled reactors. It is operated as part of the Gas-Cooled Fast Reactor programme to determine the performance of core assemblies. (U.K.)

  11. Gaseous and metallic fission product release characteristics of a modular pebble bed HTGR during loss of core cooling accidents

    International Nuclear Information System (INIS)

    A quantitative safety criteria for the high-temperature gas-cooled reactor (HTGR) is to limit the radiological consequences for a wide spectrum of accidents to a level not requiring public sheltering. This leads to reliance on passive safety characteristics for improbable loss of core cooling accidents. Models have been developed to predict the transport of metallic and gaseous fission products (FPs) through the multilayered fuel particle coatings and the graphite matrix of the core under accident conditions. Using these models, FP transport and releases were calculated for a loss of core convective cooling accident in a 250-MW(t) 3.8-W/cc pebble bed HTGR. Fission-product transport through the particle kernel and coatings, the graphite pebbles/reflectors, the reactor vessel, and the confinement were assessed. The results of this study show that the most effective barrier to fission products is the coated fuel particle. The reactor vessel and the confinement provide additional attenuation for the small amount released from the core. The small release to the environment occurs over a period of days and is so low that the safety criterion of 5 rem thyroid dose (to avoid offsite sheltering) is satisfied with a margin of more than an order of magnitude. 6 figs

  12. New steam generation system for lead-cooled fast reactors, based on steam re-circulation through ejector

    International Nuclear Information System (INIS)

    Highlights: • Innovative steam generation system for lead-cooled fast reactors secondary loop. • Water evaporation outside of vessel heated by recirculation steam in a surface exchanger. • Steam recirculation occurs through steam jet ejector feeding bayonet heat exchangers. • Improvement of safety, availability and efficiency with respect to Loeffler system (EBBSG). - Abstract: The EBBSG (External Boiling Bayonet Steam Generator) system, proposed in previous publications, offers an alternative to the classical once-through high pressure steam generators. This system exploits the combination between the Loeffler external boiling scheme and the bayonet-tube steam generator and is expected to provide advantages in terms of safety while keeping good values of cycle performance and vessel size. The main disadvantages result in the increased size of the heat exchangers with respect to once-through steam boilers and in the need of steam blowers, as envisaged under the Loeffler scheme. In the present paper, a new and more efficient system is proposed, in which the steam circulation is assured by steam-jet ejectors instead of blowers. The innovative solution, named SJ-EBBSG (Steam-Jet External Boiling Bayonet Steam Generator), is expected to provide several advantages with respect to the original scheme. In particular, the advantages envisage an increased global efficiency (+0.49% with respect to EBBSG) due to the lower power consumption of the auxiliaries and smaller size of the bayonet heat exchangers (−6.1% diameter, −7.3% length), other than increased safety and plant availability. Throughout the article, the two steam generation solutions are compared and the advantages demonstrated by calculations

  13. Core flows and heat transfer induced by inhomogeneous cooling with sub- and supercritical convection

    Science.gov (United States)

    Dietrich, W.; Hori, K.; Wicht, J.

    2016-02-01

    The amount and spatial pattern of heat extracted from cores of terrestrial planets is ultimately controlled by the thermal structure of the lower rocky mantle. Using the most common model to tackle this problem, a rapidly rotating and differentially cooled spherical shell containing an incompressible and viscous liquid is numerically investigated. To gain the physical basics, we consider a simple, equatorial symmetric perturbation of the CMB heat flux shaped as a spherical harmonic Y11 . The thermodynamic properties of the induced flows mainly depend on the degree of nonlinearity parametrised by a horizontal Rayleigh number Rah =q∗ Ra , where q∗ is the relative CMB heat flux anomaly amplitude and Ra is the Rayleigh number which controls radial buoyancy-driven convection. Depending on Rah we identify and characterise three distinctive flow regimes through their spatial patterns, heat transport and flow speed scalings: in the linear conductive regime the radial inward flow is found to be phase shifted 90° eastwards from the maximal heat flux as predicted by a linear quasi-geostrophic model for rapidly rotating spherical systems. The advective regime is characterised by an increased Rah where nonlinearities become significant, but is still subcritical to radial convection. There the upwelling is dispersed and the downwelling is compressed by the thermal advection into a spiralling jet-like structure. As Rah becomes large enough for the radial convection to set in, the jet remains identifiable on time-average and significantly alters the global heat budget in the convective regime. Our results suggest, that the boundary forcing not only introduces a net horizontal heat transport but also suppresses the convection locally to such an extent, that the net Nusselt number is reduced by up to 50%, even though the mean CMB heat flux is conserved. This also implies that a planetary core will remain hotter under a non-homogeneous CMB heat flux and is less well mixed. A

  14. Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

    2015-09-15

    Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

  15. Central mass profiles of the nearby cool-core galaxy clusters Hydra A and A478

    Science.gov (United States)

    Okabe, N.; Umetsu, K.; Tamura, T.; Fujita, Y.; Takizawa, M.; Matsushita, K.; Fukazawa, Y.; Futamase, T.; Kawaharada, M.; Miyazaki, S.; Mochizuki, Y.; Nakazawa, K.; Ohashi, T.; Ota, N.; Sasaki, T.; Sato, K.; Tam, S. I.

    2016-03-01

    We perform a weak-lensing study of the nearby cool-core galaxy clusters, Hydra A (z = 0.0538) and A478 (z = 0.0881), of which the brightest cluster galaxies (BCGs) host the powerful activities of active galactic nuclei (AGNs). For each cluster, the observed tangential shear profile is described well by either a single Navarro-Frenk-White model or a two-component model including the BCG as an unresolved point mass. For A478, we determine the BCG and its host-halo masses from a joint fit to weak-lensing and stellar photometry measurements. We find that the choice of initial mass functions (IMFs) can introduce a factor of 2 uncertainty in the BCG mass, whereas the BCG host-halo mass is constrained well by data. We perform a joint analysis of the weak-lensing and stellar kinematics data available for the Hydra A cluster, which allows us to constrain the central mass profile without assuming specific IMFs. We find that the central mass profile (r < 300 kpc) determined from the joint analysis is in excellent agreement with those from independent measurements, including dynamical masses estimated from the cold gas disc component, X-ray hydrostatic total mass estimates, and the central stellar mass estimated with the Salpeter IMF. The observed dark matter fraction around the BCG for Hydra A is found to be smaller than those predicted by adiabatic contraction models, suggesting the importance of other physical processes, such as AGN feedback and/or dissipationless mergers.

  16. Stationary light pulses and narrowband light storage in a laser-cooled ensemble loaded into a hollow-core fiber

    CERN Document Server

    Blatt, Frank; Halfmann, Thomas; Peters, Thorsten

    2016-01-01

    We report on the first observation of stationary light pulses and narrowband light storage inside a hollow-core photonic crystal fiber. Laser-cooled atoms were first loaded into the fiber core providing strong light-matter coupling. Light pulses were then stored in a collective atomic excitation using a single control laser beam. By applying a second counterpropagating control beam, a light pulse could be brought to a standstill. Our work paves the way towards the creation of strongly-correlated many-body systems with photons and applications in the field of quantum information processing.

  17. Do radio mini-halos and gas heating in cool-core clusters have a common origin?

    CERN Document Server

    Bravi, Luca; Brunetti, Gianfranco

    2015-01-01

    In this letter we present a study of the central regions of cool-core clusters hosting radio mini-halos, which are diffuse synchrotron sources extended on cluster-scales surrounding the radio-loud brightest cluster galaxy. We aim to investigate the interplay between the thermal and non-thermal components in the intra-cluster medium in order to get more insights into these radio sources, whose nature is still unclear. It has recently been proposed that turbulence plays a role for heating the gas in cool cores. By assuming that mini-halos are powered by the same turbulence, we expect that the integrated radio luminosity of mini-halos, $\

  18. Conceptual design study of Pebble Bed Type High Temperature Gas-cooled Reactor with annular core structure

    International Nuclear Information System (INIS)

    This report presents the Conceptual Design Study of Pebble Bed Type High Temperature Gas-cooled Reactor with Annular Core Structure. From this study, it is made clear that the thermal power of the Pebble Bed Type Reactor can be increased to 500MW through introducing the annular core structure without losing the inherent safe characteristics (in the coolant depressurization accident, the fuel temperature does not exceed the temperature where the fuel defect begins.) This thermal power is two times higher than the inherent safe Pebble Bed Type High temperature Gas-cooled Reactor (MHTGR) designed in West Germany. From this result, it is foreseen that the ratio of the plant cost to the reactor power is reduced and the economy of the plant operation is improved. The reactor performances e.g. fuel burnup and fuel temperature are maintained in same level of the MHTGR. (author)

  19. Loss of coolant accident analysis and evolution of emergency core cooling system for an inpile irradiation facility

    International Nuclear Information System (INIS)

    This paper deals with the Loss of Coolant Accident (LOCA) analysis of an inpile facility using RELAP4/MOD6 computer code. The present study is the culmination of a three part LOCA analysis done earlier by the authors. Blowdown analysis had been extended to include reflood part of the transient. Based on the analysis an Emergency Core Cooling System (ECCS) has been evolved. (author). 5 figs., 2 tabs

  20. 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

  1. Status of degraded core issues. Synthesis paper prepared by G. Bandini in collaboration with the NEA task group on degraded core cooling

    International Nuclear Information System (INIS)

    The in-vessel evolution of a severe accident in a nuclear reactor is characterised, generally, by core uncover and heat-up, core material oxidation and melting, molten material relocation and debris behaviour in the lower plenum up to vessel failure. The in-vessel core melt progression involves a large number of physical and chemical phenomena that may depend on the severe accident sequence and the reactor type under consideration. Core melt progression has been studied in the last twenty years through many experimental works. Since then, computer codes are being developed and validated to analyse different reactor accident sequences. The experience gained from the TMI-2 accident also constitutes an important source of data. The understanding of core degradation process is necessary to evaluate initial conditions for subsequent phases of the accident (ex-vessel and within the containment), and define accident management strategies and mitigative actions for operating and advanced reactors. This synthesis paper, prepared within the Task Group on Degraded Core Cooling (TG-DCC) of PWG2, contains a brief summary of current views on the status of degraded core issues regarding light water reactors. The in-vessel fission product release and transport issue is not addressed in this paper. The areas with remaining uncertainties and the needs for further experimental investigation and model development have been identified. The early phase of core melt progression is reasonably well understood. Remaining uncertainties may be addressed on the basis of ongoing experimental activities, e.g. on core quenching, and research programs foreseen in the near future. The late phase of core melt progression is less understood. Ongoing research programs are providing additional valuable information on corium molten pool behaviour. Confirmatory research is still required. The pool crust behaviour and material relocation into the lower plenum are the areas where additional research should

  2. Recriticality and cooling considerations of relocated molten fuel following core meltdown accident and core catcher design for PFBR

    International Nuclear Information System (INIS)

    PFBR design requires that molten fuel following a meltdown accident is relocated permanently into a coolable and sub critical configuration. Currently available information regarding the physical phenomena occurring in the course of fuel melting and relocation in a severe accident are limited to scaled down experiments involving single-pin and subassembly geometries. As shown in this note, the observed phenomena are seen to be scale dependent, making extrapolation to full-sized systems, unreliable. Therefore, one cannot count on phenomenological modeling either to rule out melt down accidents or to assess the extent of melting, if it occurs. Therefore, a core catcher for PFBR is advisable. Its size can be fixed assuming the meltdown of 7 subassemblies. The justification for the assumed extent of melting is essentially experimental. The assumed size does not lead to recriticality. In the present work, for PFBR fuel composition, (i) recriticality potential in general and (ii) that corresponding to typical design basis accident, and (iii) the coolability of the molten fuel in the core catcher are analyzed in detail. General recriticality potential of the fuel mass as a function of its mass, amount of steel that it mixes with, extent of sodium envelope, and geometrical shape it takes (spherical, hemi-spherical, and cylindrical), is investigated. Presently available design for the core catcher (for Superphenix) is considered for the PFBR and investigated. A new design for the core catcher surface is conceived and analyzed. (author)

  3. Study of the mechanisms for the emergency cooling of the core of the Radioisotope Producing Reator (RPR)

    International Nuclear Information System (INIS)

    The mechanisms for the emergency cooling of the core of the Radioisotope Producing Reactor (R.P.R.) are studied, in particular the thermal-hydraulic behaviour of the coolant after reactor shut-down. The coolant operates bd convection, and flows downward through the core passing into beel-shaped plenum that encloses the core and proceeding across the primary cooling loop. When the reactor is shut-down, the coolant flow undergoes a transient period until the steady state of natural convection is reached, after which the coolant flows upwards from the lower plenum. A plocking valve will be installed at the exit of the lower plenum, which will automatically shut in case of an accident that will involve the loss of flow in the primary circuit. The present work aims at evaluating the contribution of natural convection by natural recirculation in the core when the blocking valve is close, and via the external coolant circuit when the blocking valve is open. In particular, we study the natural self-regulating mechanisms of extraction of the heat generated by the fission product after reactor shut-down. (author)

  4. A simple model to evaluate the natural convection impact on the core transients in liquid metal cooled ads

    International Nuclear Information System (INIS)

    A simple model has been developed at ENEA Casaccia to preliminarily evaluate the primary-coolant natural convection impact on core-dynamics of an 80 Mw energy amplifier demonstration facility (EADF) fuelled by U-Pu mixed oxides and cooled by a molten lead-bismuth eutectic. The model has been already coupled with the Tieste-Minosse 'point dynamics' code elaborated at ENEA Casaccia, and in the near future will be easily coupled to the codes that are being developed at the Politecnico di Torino in the frame of the cooperation with ENEA on multi-dimensional investigations of solid fuelled ADS core dynamics. After the model formulation, some preliminary results on the primary-coolant impact on the EADF core dynamics are presented in this paper. (author)

  5. Characteristic responses of core exit thermocouples during inadequate core cooling in small break LOCA experiments conducted at Large-Scale Test Facility (LSTF) of ROSA-IV program

    International Nuclear Information System (INIS)

    Characteristic responses of core exit thermocouples (CETs) for detection of an inadequate core cooling (ICC) were experimentally studied at a large-scale plant simulator for a pressurized water reactor (PWR). The ICC conditions were established by assuming a failure or delayed actuation of high pressure injection (HPI) system. The CET responses were studied in twenty-one experiments simulating different kinds of small break loss-of-coolant accident (SBLOCA) in the PWR. It is concluded that the CETs are useful for ICC monitoring during boil-off process. An empirical equation to estimate a delay time for ICC detection is obtained for the experiments with scaled break area less than 5%. On the other hand, the ICC was not detected in 10% cold leg break test due to water falling back from the hot legs

  6. Reliability and functional testing scheme for cold circulating pumps required to cool large size fusion grade superconducting magnets and cryo-pumps

    International Nuclear Information System (INIS)

    Forced flow cooling using supercritical helium is the most preferable method due to the distinct advantages over the other cooling procedures for the superconducting magnets and cryo-pumps in fusion research devices. The flow requirements are high to fulfill the stability requirement of the magnet system during all operational modes. The flow requirements are met with cold circulation pump at 4 K level. These pumps require state of the art design due to constraints from temperature and associated process requirements with a demand of high efficiency. The future requirement of the future fusion research reactor (ITER) is foreseen as ∼ 2.7 kg/s mass flow rate with adiabatic efficiency > 70%. Against the future requirement, the maximum capacity ever built till now has a capacity of 1.2 kg/s mass flow with adiabatic efficiency ∼ 60%. Therefore, the up scaling of existing cold circulating pumps with improvement of efficiency is necessary to meet the future requirement. This paper discusses the major risks associated with cold circulating pumps and a test proposal with basic testing scheme to validate the performance. (author)

  7. Emergency cooling system for the core of a reactor pressure vessel

    International Nuclear Information System (INIS)

    In order to improve the spray distribution in an emergency cooling system for a BWR, the spray nozzles are situated vertically in bores of the pressure containment lid domed towards the inside. The distribution system is therefore situated above the lid and is supported on it. The penetrations for the incoming pipes are situated in the lid. This emergency cooling system is easy to mount and can be backfitted in existing plant. (orig./HP)

  8. Effects of cooling rate on vermicular graphite percentage in a brake drum produced by one-step cored wire injection

    Directory of Open Access Journals (Sweden)

    Yu-shuang Feng

    2015-09-01

    Full Text Available In this research, a vermicular graphite cast iron brake drum was produced by cored wire injection in a one-step method. Silica sand and low-density alumina-silicate ceramic were used as molding materials in order to investigate the effect of cooling rate on percentage of vermicular graphite and mechanical properties of the brake drum casting. Several thermocouples were inserted into the casting in the desired positions to measure the temperature change. By means of one-step cored wire injection, the two residual concentrations of Mg and RE were effectively controlled in the ranges of 0.013%-0.017% and 0.019%-0.025%, respectively, which are crucial for the production of vermicular graphite cast iron and the formation of vermicular graphite. In addition, the cooling rate had a significant effect on the vermicular graphite percentage. In the case of the silica mold brake drum casting, there was an obvious difference in the cooling rate with the wall change, leading to a change in vermicular graphite percentage from 70.8% to 90%. In the low-density alumina-silicate ceramic mold casting, no obvious change in temperature was detected by the thermocouples and the percentage of the vermicular graphite was stable at 85%. Therefore, the vermicular graphite cast iron brake drum with a better combination of mechanical properties could be obtained.

  9. Compendium of ECCS [Emergency Core Cooling Systems] research for realistic LOCA [loss-of-coolant accidents] analysis: Final report

    International Nuclear Information System (INIS)

    In the United States, Emergency Core Cooling Systems (ECCS) are required for light water reactors (LWRs) to provide cooling of the reactor core in the event of a break or leak in the reactor piping or an inadvertent opening of a valve. These accidents are called loss-of-coolant accidents (LOCA), and they range from small leaks up to a postulated full break of the largest pipe in the reactor cooling system. Federal government regulations provide that LOCA analysis be performed to show that the ECCS will maintain fuel rod cladding temperatures, cladding oxidation, and hydrogen production within certain limits. The NRC and others have completed a large body of research which investigated fuel rod behavior and LOCA/ECCS performance. It is now possible to make a realistic estimate of the ECCS performance during a LOCA and to quantify the uncertainty of this calculation. The purpose of this report is to summarize this research and to serve as a general reference for the extensive research effort that has been performed. The report: (1) summarizes the understanding of LOCA phenomena in 1974; (2) reviews experimental and analytical programs developed to address the phenomena; (3) describes the best-estimate computer codes developed by the NRC; (4) discusses the salient technical aspects of the physical phenomena and our current understanding of them; (5) discusses probabilistic risk assessment results and perspectives, and (6) evaluates the impact of research results on the ECCS regulations. 736 refs., 412 figs., 66 tabs

  10. XMM-Newton and Chandra Observations of Abell 2626: Interacting Radio Jets and Cooling Core with Jet Precession?

    CERN Document Server

    Wong, Ka-Wah; Blanton, Elizabeth L; Reiprich, Thomas H

    2008-01-01

    We present a detailed analysis of the XMM-Newton and Chandra observations of Abell 2626 focused on the X-ray and radio interactions. Within the region of the radio mini-halo (~70 kpc), there are substructures which are probably produced by the central radio source and the cooling core. We find that there is no obvious correlation between the radio bars and the X-ray image. The morphology of Abell 2626 is more complex than that of the standard X-ray radio bubbles seen in other cool core clusters. Thus, Abell 2626 provides a challenge to models for the cooling flow -- radio source interaction. We identified two soft X-ray (0.3--2 keV) peaks with the two central cD nuclei; one of them has an associated hard X-ray (2--10 keV) point source. We suggest that the two symmetric radio bars can be explained by two precessing jets ejected from an AGN. Beyond the central regions, we find two extended X-ray sources to the southwest and northeast of the cluster center which are apparently associated with merging subclusters...

  11. The SKA view of cool-core clusters: evolution of radio mini-halos and AGN feedback

    CERN Document Server

    Gitti, Myriam; Brunetti, Gianfranco; Cassano, Rossella; Dallacasa, Daniele; Edge, Alastair; Ettori, Stefano; Feretti, Luigina; Ferrari, Chiara; Giacintucci, Simona; Giovannini, Gabriele; Hogan, Michael; Venturi, Tiziana

    2014-01-01

    In about 70% of the population of relaxed, cool-core galaxy clusters, the brightest cluster galaxy (BCG) is radio loud, showing non-thermal radio jets and lobes ejected by the central active galactic nucleus (AGN). In recent years such relativistic plasma has been unambiguously shown to interact with the surrounding thermal intra-cluster medium (ICM) thanks to spectacular images where the lobe radio emission is observed to fill the cavities in the X-ray-emitting gas. This `radio feedback' phenomenon is widespread and is critical to understand the physics of the inner regions of galaxy clusters and the properties of the central BCG. At the same time, mechanically-powerful AGN are likely to drive turbulence in the central ICM which may also play a role for the origin of non-thermal emission on cluster-scales. Diffuse non-thermal emission has been observed in a number of cool-core clusters in the form of a radio mini-halo surrounding the radio-loud BCG on scales comparable to that of the cooling region. Large mi...

  12. Development of simple method to incorporate out-of-core cooling effect on thorium conversion in multi-pass fueled reactor and investigation on characteristics of the effect

    International Nuclear Information System (INIS)

    Highlights: • I proposed simple method based on analytical approach. • I applied this method to MVP-BURN calculations without a code modification. • I estimated conversion efficiencies based on this method. • I compared the conversion efficiencies of MSBR and PBMR. - Abstract: Development of a simple method to incorporate the out-of-core cooling effect on the thorium conversion in multi-pass fueled reactors and investigation on characteristics of the effect have been undertaken. For multi-pass fueled reactors, such as Molten Salt Breeder Reactor (MSBR) and Pebble-Bed Modular Reactor (PBMR), fuel moves in the core and exits from the core. The produced nuclides also decay out of the core, which should be considered for core characteristics when needed. In the present study, 233Pa is selected to evaluate the thorium conversion accurately. To take the effect into account, in the present study, an effective decay constant is proposed to make equilibrium concentration of 233Pa without out-of-core cooling equal to that with out-of-core cooling. With the effective decay constant, the out-of-core cooling effect can be incorporated even with the code system using macroscopic cross sections generated by cell burn-up calculations without any code modification. In addition, the characteristics of out-of-core cooling effect for the thorium conversion are evaluated for thorium fueled reactors MSBR and PBMR. It is concluded that the out-of-core cooling effect is suitable for MSBR to enhance thorium conversion because of the fast flow rate of fuel salt. On the other hand, the effect is neither important nor realistic to employ for PBMR because the in-core residence time of approximately 100 days is longer than the half-life of 233Pa of 27.0 days, and the effect cannot improve the conversion ratio significantly

  13. AGN driven perturbations in the intra-cluster medium of cool core cluster ZwCl 2701

    CERN Document Server

    Vagshette, Nilkanth D; Naik, Sachindra; Patil, Madhav K

    2016-01-01

    We present the results obtained from a total of 123 ks X-ray (Chandra) and 8 hrs of 1.4 GHz radio (Giant Metrewave Radio Telescope - GMRT) observations of the cool core cluster ZwCl 2701 (z = 0.214). These observations of ZwCl 2701 showed the presence of an extensive pair of ellipsoidal cavities along the East and West directions within the central region < 20 kpc. Detection of bright rims around the cavities suggested that the radio lobes displaced X-ray emitting hot gas forming shell-like structures. The total cavity power (mechanical power) that directly heated the surrounding gas and cooling luminosity of the cluster were estimated to be ~2.27 x 10^{45} erg\\s and 3.5 x 10^{44} erg\\s, respectively. Comparable values of cavity power and cooling luminosity of ZwCL 2701 suggested that the mechanical power of the AGN outburst is large enough to balance the radiative cooling in the system. The star formation rate derived from the H_alpha luminosity was found to be ~0.60 M_sun yr^{-1} which is about three ord...

  14. Performance of the lift-pump with the lead-bismuth cooled fast reactor. Experimental study on bubble distribution and circulation flow rate

    International Nuclear Information System (INIS)

    Recently, the utilization of the lift pump is examined in a small reactor of the lead-bismuth eutectic cooling. Then, the experiments concerning about void behavior and performance of the lift pump in three kinds of risers (1124mm in height and inside diameters φ69.3mm, φ106.3mm, φ155.2mm) were performed by using lead bismuth eutectic. The main results are as follows: (1) The local void fraction varies in horizontal plane in case of the big diameter riser. (2) The lead-bismuth circulating flow rate evaluated by a present design method becomes lower than that of experiments in case of medium and small diameter risers. This design method can be used as an outline evaluating function for these cases, considering the evaluation accuracy of the pressure loss of the test section in the calculation. (3) In the big diameter riser, the present design method excessively evaluates the lead-bismuth circulating flow rate. It thought that the circulation head will not occur in the experiments such as a results of the present design method because the void rises biasing in horizontal plane in case of big diameter riser though the present method is one dimensional model. It is better to utilize a separator which can divides the riser into about 10cm diameter flow path and the void is fed uniformly distributed to each paths to obtain appropriate circulation head. (author)

  15. A Core Design Approach Aimed at the Sustainability and Intrinsic Safety of the European Lead-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Among the Generation-IV fast reactor technologies, a Lead-cooled Fast Reactor concept is currently under development in Europe as a potential candidate for the deployment, to meet long-term objectives of European energy policies. Within the Lead-cooled European Advanced DEmonstration Reactor (LEADER) project, co-financed by the European Union within the 7th EURATOM Framework Programme, the conceptual design of the reference Generation-IV European LFR (ELFR) industrial plant was developed, benefiting from and further optimizing the concept put forward during the ELSY 6th EURATOM Framework Programme project. In order to embed in the design the safety and sustainability goals in the most effective way, an innovative, dedicated design approach was developed and applied to the design of the ELFR fuel pins, fuel assemblies and core. This new approach, together with the main analysis results supporting the design of the reference ELFR configuration, are presented and discussed in detail. (author)

  16. Annular core for modular high temperature gas-cooled reactor (MHTGR)

    International Nuclear Information System (INIS)

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40 % greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8 % enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above. (author)

  17. SEARCHING FOR COOLING SIGNATURES IN STRONG LENSING GALAXY CLUSTERS: EVIDENCE AGAINST BARYONS SHAPING THE MATTER DISTRIBUTION IN CLUSTER CORES

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, Peter K. [Astronomy Department, University of California, Berkeley, B-20 Hearst Field Annex 3411, Berkeley, CA 94720-3411 (United States); Bayliss, Matthew B. [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); McDonald, Michael [Massachusetts Institute of Technology, Kavli Institute for Astrophysics and Space Research, 77 Massachusetts Avenue 37-287, Cambridge, MA 02139 (United States); Dahle, Hakon [Institute of Theoretical Astrophysics, University of Oslo, P.O. Box 1029, Blindern, NO-0315 Oslo (Norway); Gladders, Michael D. [Department of Astronomy and Astrophysics, University of Chicago, 5640 South Ellis Avenue, Chicago, IL 60637 (United States); Sharon, Keren [Department of Astronomy, University of Michigan, 500 Church Street, Ann Arbor, MI 48109-1042 (United States); Mushotzky, Richard, E-mail: pblanchard@fas.harvard.edu [Astronomy Department, University of Maryland, College Park, MD 20742 (United States)

    2013-07-20

    The process by which the mass density profile of certain galaxy clusters becomes centrally concentrated enough to produce high strong lensing (SL) cross-sections is not well understood. It has been suggested that the baryonic condensation of the intracluster medium (ICM) due to cooling may drag dark matter to the cores and thus steepen the profile. In this work, we search for evidence of ongoing ICM cooling in the first large, well-defined sample of SL selected galaxy clusters in the range 0.1 < z < 0.6. Based on known correlations between the ICM cooling rate and both optical emission line luminosity and star formation, we measure, for a sample of 89 SL clusters, the fraction of clusters that have [O II]{lambda}{lambda}3727 emission in their brightest cluster galaxy (BCG). We find that the fraction of line-emitting BCGs is constant as a function of redshift for z > 0.2 and shows no statistically significant deviation from the total cluster population. Specific star formation rates, as traced by the strength of the 4000 A break, D{sub 4000}, are also consistent with the general cluster population. Finally, we use optical imaging of the SL clusters to measure the angular separation, R{sub arc}, between the arc and the center of mass of each lensing cluster in our sample and test for evidence of changing [O II] emission and D{sub 4000} as a function of R{sub arc}, a proxy observable for SL cross-sections. D{sub 4000} is constant with all values of R{sub arc}, and the [O II] emission fractions show no dependence on R{sub arc} for R{sub arc} > 10'' and only very marginal evidence of increased weak [O II] emission for systems with R{sub arc} < 10''. These results argue against the ability of baryonic cooling associated with cool core activity in the cores of galaxy clusters to strongly modify the underlying dark matter potential, leading to an increase in SL cross-sections.

  18. Specialists' meeting on instrumentation for supervision of core cooling in FBRs, Kalpakkam, India, 12-15 December 1989

    International Nuclear Information System (INIS)

    The purpose of the meeting was to provide a forum to discuss instrumentation provisions required for the assurance of core cooling in all operating conditions covering needs for both global and local supervision. The presentations by the participants were divided into four topical sessions: national position papers, operating experience, advanced measurement techniques, signal processing techniques. Twenty specialists from six countries and the IAEA took part in the meeting. Fifteen papers were presented. A separate abstract was prepared for each of these papers. After the formal sessions were completed, a final discussion session was held and general conclusions and recommendations were reached. Refs, figs and tabs

  19. 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)

  20. Development of an emergency core cooling system for the converted IEA-R1m research reactor

    International Nuclear Information System (INIS)

    This present work describes the development program carried out in the design and construction of the Emergency Core Cooling System for the IEA-R1m Research Reactor, including the system design, the experiments performed to validate the design, manufacturing, installation and commissioning. The experiments were performed in two phases. In the first phase, the spray flow rate and distribution were measured, using a full scale mock-up of the entire core, to establish the spray header geometry and specifications. In the second phase, a test section was fitted with electrically heated plates to simulate the fuel plates. Temperature measurements were carried out to demonstrate the effectiveness of the system to keep the temperatures below the limiting value. The experimental results were shown to the licensing authorities during the certification process. The main difficulties during the system assembly are also described. (author)

  1. Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

    International Nuclear Information System (INIS)

    Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.

  2. Thermohydraulic characteristics analysis of natural convective cooling mode on the steady state condition of upgraded JRR-3 core, using COOLOD-N code

    International Nuclear Information System (INIS)

    This report describes the results of the steady state thermohydraulic analysis of upgraded JRR-3 core under natural convective cooling mode, using COOLOD-N code. In the code, function to calculate flow-rate under natural convective cooling mode, and a heat transfer package have been newly added to the COOLOD code which has been developed in JAERI. And this report describes outline of the COOLOD-N code. The results of analysis show that the thermohydraulics of upgraded JRR-3 core, under natural convective cooling mode have enough margine to ONB temperature, DNB heat flux and occurance of blisters in fuel meats, which are design criterion of upgraded JRR-3. (author)

  3. XMM-Newton Observations of A133: A Weak Shock Passing through the Cool Core

    CERN Document Server

    Fujita, Yutaka; Reiprich, Thomas H; Andernach, H; Ehle, M; Murgia, M; Rudnick, L; Slee, O B

    2004-01-01

    We use XMM-Newton observations of the cluster of galaxies A133 to study the X-ray spectrum of the intracluster medium (ICM). We find a cold front to the southeast of the cluster core. From the pressure profile near the cold front, we derive an upper limit to the velocity of the core relative to the rest of the cluster of ~1.5\\mu G.

  4. Core design of heavy water cooled thorium breeder reactor with negative void reactivity and improved breeding performance

    International Nuclear Information System (INIS)

    A core of heavy water cooled thorium breeder reactor that produces 3.5 GWt energy using Th-233U oxide fuel has been studied to depict a concrete design specification. In order to improve the breeding performance compared to that of our previous study, one of key parameters in core design: moderator to fuel volume ratio (MFR) is re-surveyed. By reducing MFR from 1.0 to 0.6, the swing of keff during a cycle is considerably flattened, keeping negative void coefficient. The batch number is 3 and the refueling scheme employs out-in method to limit the radial power peaking factor less than 1.3. Due to efficient internal conversion, the reactivity of the core slightly increases with burnup, so that the cycle length is extended up to 1,300 days. Consequently, high averaged burnup of 80 GWd/t and breeding ratio of 1.07 at middle of cycle is achieved without any blankets. The number of control rods made of B4C is 19 and the total reactivity worth is -6.5% dk/k. The present core uses Zircaloy-4 as cladding material, the fast neutron fluence at EOC (End Of Cycle), however, exceeds its limit due to hard spectrum and long cycle length. As a part of future study, design will be further explored considering cladding integrity. (authors)

  5. Analysis and Determination of Temperature in the B Ring of the KartiniReactor Core Primary Cooling Water for 250 kW

    International Nuclear Information System (INIS)

    Analysis and determination of temperature in the B ring of primarycooling water in the Kartini reactor core for 250 kW of power level has beendone. The Instrumented Fuel Element (IFE) is used for measurement oftemperature with varying of power from 10 kW to 100 kW. Heat transfer ishappened from the fuel to the core primary cooling water as a function ofpower level. The calculation for determining of temperature of primarycooling water indicate that for operation 250 kW power level the temperatureof the core primary cooling water is Tp = 117.478 oC. (author)

  6. 食品工业循环冷却水中藻类的去除%Removal of Algae from Circulated Cooling Water in Food Industry

    Institute of Scientific and Technical Information of China (English)

    郑必胜; 蔡妙颜; 郭祀远; 李琳; 张智平

    2001-01-01

    The application of gas floatation to the removing of algae from circulated cooling water in food industry was studied. The proper flocculant was selected and influence of some factors, such as concentration of flocculant, pH and the characteristic of air bubble, on the removing of algae was discussed. By the way, the optimum treatment condition was put forward. It is verified that gas floatation is the best method for the removing of algae from the circulated cooling water.%研究采用絮凝气浮法除去食品工业循环冷却水中的藻类,选择适当的絮凝剂并探讨了絮凝剂用量、pH、气泡特性等因素对藻类除去效果的影响,提出了较合理处理条件。结果表明,絮凝气浮法是处理食品工业循环冷却水的较佳方法。

  7. Flow distribution experimental study on the emergency core cooling system of the IEA-R1m - IPEN-CNEN/SP - Brazil

    International Nuclear Information System (INIS)

    This paper presents a brief description of Emergency Core Cooling System designed by the IEA-R1m Reactor and the experimental results of flow distribution over the core. Several parameters were evaluated, such as: relative position of spray header to the reactor core; type and quantity of spray nozzles; spray nozzles position on spray header; and total spray flow. The main conclusions are presented. (author)

  8. Core melt cooling inside the vessel through external cavity two-phase flow

    International Nuclear Information System (INIS)

    The possibility of containing the corium inside the vessel by an external cooling water system has been investigated by studies based on two kinds of thermal-hydraulic computations: calculations with the TRIO-VF (Finite volume version) code of the behavior of a corium pool inside the lower head of the vessel in order to assess the heat flux distribution on the wall; calculations of the external cooling of the vessel with the finite element two-phase flow TRIO-GENEPI code. The cases of a constant heat flux and the heat flux distribution supplied by the TRIO-VF computations are examined. Results show a stratified boiling flow along the heating lower head (case 1) and along its upper part (case 2). A dryout is predicted in the case 1 only. 14 figs., 1 tab., 10 refs

  9. Test of a cryogenic set-up for a 10 meter long liquid nitrogen cooled superconducting power cable

    DEFF Research Database (Denmark)

    Træholt, Chresten; Rasmussen, Carsten; Kühle (fratrådt), Anders Van Der Aa; Olsen, Søren Krüger; Jensen, Kim Høj; Tønnesen, Ole

    2000-01-01

    High temperature superconducting power cables may be cooled by a forced flow of sub-cooled liquid nitrogen. One way to do this is to circulate the liquid nitrogen (LN2) by means of a mechanical pump through the core of the cable and through a sub-cooler.Besides the cooling station, the cryogenics...

  10. Lumped parameter analysis of Pb-Bi cooled fast reactor PEACER core using MATRA

    International Nuclear Information System (INIS)

    PEACER core is designed to produce 1560MW of thermal output with electric output of 550MW. PEACER uses B4C control rods and lead-bismuth (Pb-Bi) coolant in the primary system. This work examines the Pb-Bi coolant behavior along the PEACER fuel channels and to check on whether the given heat flux profiles and geometrical arrangement of the fuel rods yield reasonable fluid dynamic distribution under nominal operation resorting to a subchannel approach using MATRA. MATRA is a thermal hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in the fuel rod bundle during steady-state and transient conditions. The calculational result revealed that the input data based on the current design of PEACER core yielded reasonable results mostly satisfying the thermal design limits. The calculation results, however, indicated a potential for fuel damage in the hottest assembly of the core. This was found to be mainly due to excessively conservative assumptions made in generating the input conditions. Work is underway to apply physically-based conditions of the PEACER core and more reliable rod-to-coolant heat transfer correlations. (author)

  11. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  12. Analytical and CFD investigation of ex-core cooling of the nuclear fuel rod bundle in a water pool

    International Nuclear Information System (INIS)

    The efficiency of ex-core cooling of nuclear fuel assemblies under decay heat generation is influenced by many conditions, among them being coolant flow rate, position of fuel assemblies in a water pool, and position of coolant inlets and outlets. A combination of unacceptable thermal-hydraulic conditions occurred at the Nuclear Power Plant PAKS in Hungary in April 2003, during the process of nuclear fuel assembly chemical cleaning in a specially designed tank. The cooling of the nuclear fuel rod bundles in the tank was not efficient under low coolant flow rates through the cleaning tank, and after several hours the boiling of cooling water occurred with subsequent dry-out of nuclear fuel rod bundles. The thermal-hydraulic conditions in the cleaning tank that led to the unexpected event are analysed both analytically and with a CFD approach for idealized conditions of one nuclear fuel rod bundle with the bottom by-pass opening. The analytical analysis is based on a pressure balance of low Reynolds number upward water coolant flow through the bundle, downward water flow in the pool around the bundle, flow across the by-pass opening and outlet flow from the cleaning vessel. The transient CFD simulations are performed in order to demonstrate multidimensional effects of the event. The water density dependence on the temperature is taken into account in both analytical and CFD investigation, as the dominant effect that influences the buoyancy forces between the water flow streams inside and outside the vertically positioned bundle in the water pool. The influence of the bundle bottom by-pass area on the water pool thermal-hydraulic conditions and on the efficiency of the nuclear fuel rods cooling is analysed. Both analytical and CFD results show that the continuous cooling of the fuel rods can not be achieved for higher values of the bundle bottom by-pass areas. The averaged coolant temperature in the water pool outside the bundle becomes higher than the average

  13. High-dose diazepam facilitates core cooling during cold saline infusion in healthy volunteers.

    Science.gov (United States)

    Hostler, David; Northington, William E; Callaway, Clifton W

    2009-08-01

    Studies have suggested that inducing mild hypothermia improves neurologic outcomes after traumatic brain injury, major stroke, cardiac arrest, or exertional heat illness. While infusion of cold normal saline is a simple and inexpensive method for reducing core temperature, human cold-defense mechanisms potentially make this route stressful or ineffective. We hypothesized that intravenous administration of diazepam during a rapid infusion of 30 mL.kg-1 of cold (4 degrees C) 0.9% saline to healthy subjects would be more comfortable and reduce core body temperature more than the administration of cold saline alone. Fifteen subjects received rapidly infused cold (4 degrees C) 0.9% saline. Subjects were randomly assigned to receive, intravenously, 20 mg diazepam (HIGH), 10 mg diazepam (LOW), or placebo (CON). Main outcomes were core temperature, skin temperature, and oxygen consumption. Data for the main outcomes were analyzed with generalized estimating equations to identify differences in group, time, or a group x time interaction. Core temperature decreased in all groups (CON, 1.0 +/- 0.2 degrees C; LOW, 1.4 +/- 0.2 degrees C; HIGH, 1.5 +/- 0.2 degrees C), while skin temperature was unchanged. Mean (95% CI) oxygen consumption was 315.3 (253.8, 376.9) mL.kg-1.min-1 in the CON group, 317.9 (275.5, 360.3) in the LOW group, and 226.1 (216.4, 235.9) in the HIGH group. Significant time and group x time interaction was observed for core temperature and oxygen consumption (p < 0.001). Administration of high-dose diazepam resulted in decreased oxygen consumption during cold saline infusion, suggesting that 20 mg of intravenous diazepam may reduce the shivering threshold without compromising respiratory or cardiovascular function. PMID:19767791

  14. Overview on stability of natural-circulation-cooled boiling water reactors during start-up. An experimental and modeling analysis

    International Nuclear Information System (INIS)

    This paper provides an overview on numerical and experimental work focused on flashing-induced instabilities. These instabilities may occur in natural circulation two-phase systems when operated at low pressure and low power. Therefore they are of special interest for the start-up phase of natural circulation Boiling Water Reactors. The work presented in this paper has been performed within the framework of the NACUSP project (European-Union Fifth Framework Program). Experiments were carried out on a steam/water natural circulation loop (CIRCUS), built at the Delft University of Technology. Information was gained on the characteristics of the flow oscillations and on the void fraction production during flashing in stationary and transient conditions. A 3-D flow-pattern visualization was achieved by means of advanced instrumentation, namely wire-mesh sensors. On the basis of the experimental results, an assessment of existing drift-flux models was performed for flashing flow. The most suitable drift-flux model was implemented in the 4-equations two-phase model FLOCAL, developed at the Forschungszentrum Rossendorf (FZR, Germany). The model allows for the liquid and steam to be in thermal non-equilibrium and, via drift-flux models, to have different velocities. A detail comparison between simulations and experiments is reported. (author)

  15. Cores and fuel cycle of the perspective fast sodium-cooled reactor

    International Nuclear Information System (INIS)

    Full text: A perspective sodium reactor is under development in Russia nowadays. Initially, power level of 1800 MW (el.) was considered for this reactor. However, owing to many reasons, in particular, for transportability of the main plant by railway, the reactor power was later reduced to 1200 MW (el.). At the same time the base of the concept for the choice of the core parameters remained the same as for the 1800 MW power, including the following: - low core specific power resulting in a decrease of the fuel lifetime and, consequently, a smaller annual consumption of fuel elements; - enhancement of inherent self-protection: ensuring the sodium void reactivity effect (SVR) close to zero and a minimum reactivity margin for burnup; - ensuring the reactor operation in different patterns of the closed fuel cycle organization: the use of plutonium from thermal reactor with and without MA for the first loading, recycling the own plutonium with/without breeding, burnup of own MA, etc. Basic characteristics of the core of BN-1200 reactor approved for the current phase of designing have been reported. The principle of layout with upper sodium plenum, like the BN-800 reactor type is preserved in the approved variant of the core for ensuring the SVR close to zero.It is an important feature of the core layout that fuel of one enrichment level is used. This approach simplifies the technological process of manufacturing the fuel elements and fuel subassemblies (SA) and the process of SA handling at NPP. The Rules of nuclear safety (PBYaRU AS) were altered in 2008 in Russia, the requirement of negative reactivity coefficient from the volume fraction of coolant, i.e., the SVR close to zero, was withdrawn. This allows an extension of the area of optimal values for the core parameters, in particular, an extension of the core height and introduction of the top axial breeding blanket. However, inspite of a reduced strictness of regulatory requirements, the question of changeover to

  16. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    International Nuclear Information System (INIS)

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  17. Neutronics aspects associated to the prevention and mitigation of severe accidents in sodium cooled reactor cores

    International Nuclear Information System (INIS)

    Among all the types of accidents to be considered for the safety licensing of a plant, some have a very low probability of occurrence but might have very important consequences: the severe accidents or Hypothetical Core Disruptive Accidents (HCDA). The studies on the scenario of these accidents are performed in parallel to the prevention studies. In this PhD report, two representative safety cases are studied: the Unprotected Loss Of Flow (ULOF) and the Total Instantaneous Blockage (TIB). The objectives are to understand what causes the reactivity increase during these accidents and to find means to reduce the energetic release of the scenario (ULOF) or to find ways to trigger the core prior to the propagation of the accident (TIB). At first, the accidents are studied in static calculations with the ERANOS code system. The accidents are divided into several steps and the reactivity insertions at each step are explained. This study shows the importance of the removal of the structures as well as of the radial leakage changes during the core slumping-down. The study also gives the amounts of fuel to be ejected or of absorber to be injected in both accidents. These values give tracks to the following more accurate studies, the transient studies. The transient studies were performed with the SIMMER code system, coupling thermo-hydraulics and neutronics. SIMMER data and algorithms have been improved so as to better predict ERANOS results (former discrepancies were up to 1.5$). The SIMMER reactivity calculation is improved by 0.8$ with variations of reactivity due to the motion of materials correctly predicted. A new algorithm for the β-effective was implemented in SIMMER so as to be more accurate and easier to manage. SIMMER is then used to calculate the secondary phase of the ULOF, while the primary phase is calculated with ERANOS thanks to some assumptions. The assumptions are very much based on the fact that the movement of materials stops whenever the energy

  18. INTENSIFICATION OF COOLING THE POLYMER OR RUBBER ISOLATION APPLIED TO THE CORES OF CABLE PRODUCTS

    OpenAIRE

    Мікульонок, Ігор Олегович; Сокольський, Олександр Леонідович; Соколенко, В. В.

    2015-01-01

    In manufacture of cable production with insulation on the basis of high-molecular substances necessary productivity of technological lines is usually reached by application of a high-efficiency extrusion method.During formation of electric insulation from polymers and rubber mixes the temperature of an insulating cover of current-carrying cores changes from temperature of formation of polymer or vulcanization of a rubber mix to temperature in receiving device of a technological line for impos...

  19. Hydraulic characteristics of the N Reactor core and reactor cooling system

    International Nuclear Information System (INIS)

    In conjunction with the NUSAR program, the need was recognized for well substantiated pressure drop correlations for the N Reactor core to support in-depth safety analysis consistent with currently-available technology. Additionally, it was considered desirable to reconfirm the hydraulic characteristics of the reactor coolant system in the light of improved understanding of the hydraulic features of the current reactor fuel loading. The report summarizes the results of laboratory tests and analysis accomplished to meet the above objectives

  20. Lumped parameter analysis of Pb-Bi cooled fast reactor PEACER core using MATRA

    International Nuclear Information System (INIS)

    The PEACER (Proliferation-resistant Environment-friendly Accident-tolerant Continuable energy Economical Reactor) system is under study to transmute long-lived fission products and actinides as well as to produce electricity. It is important to keep the temperature of the reactor core structures under certain criteria in order to prevent damage of fuel materials which can advance to severe situations such as radiation leakage, and even meltdown of the fuel. This study intends to examine the liquid metal coolant behavior along the PEACER fuel channels and to find out whether the given heat flux profiles and geometrical arrangement of the fuel rods yield reasonable flow distribution during the nominal operation by using subchannel approach. The subchannel analysis of the fuel assembly under nominal operation condition was performed using MATRA (Multi-channel Analyzer for Transient and steady-state in Rod Arrays). The result showed that the input data based on the current design of the PEACER core yielded reliable results satisfying the thermla and mechanical design limits. Typical results obtained include the hydrodynamic conditions of Pb-Bi in subchannels and the thermodynamic states of the core structures. (author)

  1. Core Design Studies for TRU Transmutation in a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The objectives of this research project is (1) to develop the conceptual core designs for TRU transmutation covering a large variation in power level and conversion ratio and (2) to perform relevant verification and validation analyses through the analyses of fast critical experimental assemblies. An homogeneous and detailed heterogeneous models of metal fueled critical assemblies, BFS-73-1, BFS-75-1, and BFS-55-1, were produced from this study through a review of the critical experiments. Based on these models, BFS critical assemblies were analyzed by a fast reactor analysis code system (TRANSX/ TWODANT/DIF3D) with different evaluated nuclear data files including ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, JENDL-AC2008 in addition to ENDF/B-VI.6. A study of library difference on computational results by both a conventional diffusion method and a Monte-Carlo transport method has been carried out with those models. In addition to the analysis by the design code, Monte-Carlo high fidelity simulation was carried out to support the diffusion solution, mainly an effect of unit fuel cell heterogeneity. BFS and ZPPR critical assemblies were analyzed by both KAERI and ANL systems and the results of the analyses were reviewed by the other side. This improve the reliability of the results of both institutes. For the effective TRU transmutation, the conceptual core design was performed under core power ranged from 1,500MWt to 4,500MWt and found that there is no appreciable degradation in performance or reactivity coefficients for the core power level up to 1,800 MWe and confirmed the possibility of the large scaled transmutation reactor. Even at each pre-determined power level, performance parameters, reactivity coefficients and its implication on the safety analysis can be different when a target TRU conversion ratio changes. In order to address this aspect of design, a variation study of TRU conversion ratio change was covered. Three ATWS events such as UTOP, ULOF and ULOHS are

  2. Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

    International Nuclear Information System (INIS)

    Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

  3. Start-up fuel and power flattening of sodium-cooled candle core

    International Nuclear Information System (INIS)

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

  4. The prediction of two phase mixture level and cooling conditions during a partial core uncovery

    International Nuclear Information System (INIS)

    A model for prediction of the reactor core two phase mixture level, as a function of the downcomer level, has been developed. This model assumes quasi-stationary conditions and is applicable at decay heat levels. Another developed model describes the core uncovery process when no make up water is available. In a third model the heat transfer in the uncovered part of the rod bundle is predicted and the rod temperature, as well as the steam superheat temperature, is calculated as a function of time and elevation. This model can be applied for rod temperatures well above 1200 degrees C. These models have been combined and transferred to a computer code and quantities calculated by this such as axial void distribution, two phase level, and rod temperatures have been compared with test data. Comparisons of two phase level and void distributions show good agreement between test data and calculations, within the observed pressure range of 13 to 70 bar. Comparisons of rod temperatures show that they are underpredicted. This is mainly due to the assumption in the model that the channel wall has the same temperature as the fuel rods. This assumption is not valid for the tests compared with in appendix A, in which the channel box is very oversized compared to reactor conditions. In appendix B comparisons with ASEA-ATOM, DRAGON code calculations show that the present model yields a good representation of the core heat up for typical reactor conditions. A comparison with experimental temperature data from the NEPTUN test facility show generally good agreement for all tests compared with in appendix C. (author)

  5. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    Energy Technology Data Exchange (ETDEWEB)

    Rubio, Rafael, E-mail: rrubio@iberdrola.es [Iberdrola Generación Nuclear S.A., Madrid (Spain); Jimenez, Gonzalo [Universidad Politécnica de Madrid (Spain)

    2014-08-15

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available.

  6. Potential issues related to emergency core cooling system strainers performance at boiling water reactors: Application to Cofrentes NPP (Spain)

    International Nuclear Information System (INIS)

    Highlights: • Design of the ECCS strainers introduced a reasonable margin for operation in BWRs. • Studies are addressing the effects of post-LOCA debris on ECCS in Cofrentes NPP. • The head loss due is at most half of the limited head loss for the ECCS strainer. • The NPSH required is at least three times lower than the NPSH available. - Abstract: From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at boiling water reactors (BWR). In Spain the regulatory body (Consejo de Seguridad Nuclear, CSN) adopted the USA regulation and Cofrentes NPP installed new strainers with a considerable bigger size than the old strainers. The nuclear industry conducted significant and extensive research, guidance development, testing, reviews, and hardware and procedure changes during the 90s to resolve the issues related to debris blockage of BWR strainers. In 2001 the NRC and CSN closed the Bulletins. Thereafter, the strainers issues were moved to the PWR reactors. In 2004 the NRC issued a Generic Letter (GL). It requested the resolution of several effects which were not noted in the past. The GL regarded to be resolved by the PWR reactors but the NRC in USA and the CSN in Spain have requested that the BWR reactors investigate differences between the methodologies used by the BWRs and PWRs. The developments and improvements done for Cofrentes NPP are detailed. Studies for this plant show that the head loss due to the considered debris is at most half of the limited head loss for the ECCS strainer and the NPSH (Net Positive Suction Head) required for the ECCS pumps is at least three times lower than the NPSH available

  7. Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

    International Nuclear Information System (INIS)

    Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement

  8. Searching for Cooling Signatures in Strong Lensing Galaxy Clusters: Evidence Against Baryons Shaping the Matter Distribution in Cluster Cores

    CERN Document Server

    Blanchard, Peter K; McDonald, Michael; Dahle, Hakon; Gladders, Michael D; Sharon, Keren; Mushotzky, Richard

    2013-01-01

    The process by which the mass density profile of certain galaxy clusters becomes centrally concentrated enough to produce high strong lensing (SL) cross-sections is not well understood. It has been suggested that the baryonic condensation of the intra-cluster medium (ICM) due to cooling may drag dark matter to the cores and thus steepen the profile. In this work, we search for evidence of ongoing ICM cooling in the first large, well-defined sample of strong lensing selected galaxy clusters in the range 0.1 0.2 and shows no statistically significant deviation from the total cluster population. Specific star formation rates, as traced by the strength of the 4000 angstrom break, D_4000, are also consistent with the general cluster population. Finally, we use optical imaging of the SL clusters to measure the angular separation, R_arc, between the arc and the center of mass of each lensing cluster in our sample and test for evidence of changing [OII] emission and D_4000 as a function of R_arc, a proxy observable ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-31

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

  10. Restrictive effect of ascending steam on falling water during top spray emergency core cooling

    International Nuclear Information System (INIS)

    Water spraying experiments were conducted to find out a flow rate of falling water overcoming ascending steam during top spray emergency cooling with an 8 x 8 type simulated fuel rod bundle of real size. The bundle consisted of 64 rods, each with a diameter of 12.5 mm, arranged in the form of square lattice with a pitch of 16.3 mm. In the experiments the simulated fuel rods were not heated. Instead, steam was injected into the lower plenum vessel simulating bundle-generated steam. As the results, (1) a criterion was proposed to determine the region where the restrictive effect of ascending steam on falling water appears, considering the decrease of a flow rate of ascending steam due to condensation by a spray of subcooled water, (2) the restrictive effect was independent of water head on the upper tie plate and water injection methods, and (3) an analytical model based on the pressure balance at the upper tie plate was proposed to calculate a flow rate of falling water overcoming ascending steam. (author)

  11. Far Ultraviolet Morphology of Star Forming Filaments in Cool Core Brightest Cluster Galaxies

    CERN Document Server

    Tremblay, Grant R; Baum, Stefi A; Mittal, Rupal; McDonald, Michael; Combes, Françoise; Li, Yuan; McNamara, Brian; Bremer, Malcolm N; Clarke, Tracy E; Donahue, Megan; Edge, Alastair C; Fabian, Andrew C; Hamer, Stephen L; Hogan, Michael T; Oonk, Raymond; Quillen, Alice C; Sanders, Jeremy S; Salomé, Philippe; Voit, G Mark

    2015-01-01

    We present a multiwavelength morphological analysis of star forming clouds and filaments in the central ($ 5$ \\Msol) stars reveals filamentary and clumpy morphologies, which we quantify by means of structural indices. The FUV data are compared with X-ray, Ly$\\alpha$, narrowband H$\\alpha$, broadband optical/IR, and radio maps, providing a high spatial resolution atlas of star formation locales relative to the ambient hot ($\\sim10^{7-8}$ K) and warm ionised ($\\sim 10^4$ K) gas phases, as well as the old stellar population and radio-bright AGN outflows. Nearly half of the sample possesses kpc-scale filaments that, in projection, extend toward and around radio lobes and/or X-ray cavities. These filaments may have been uplifted by the propagating jet or buoyant X-ray bubble, or may have formed {\\it in situ} by cloud collapse at the interface of a radio lobe or rapid cooling in a cavity's compressed shell. The morphological diversity of nearly the entire FUV sample is reproduced by recent hydrodynamical simulations...

  12. The reactivity effects of steam ingress into the core of gas-cooled fast reactors

    International Nuclear Information System (INIS)

    Steam ingress reactivity effect is caused by steam ingress into the core of GCFRs through coolant channels from the secondary coolant system during a hypothetical accident. This reactivity effect in GCFR can be considered one of the most important safety-related physics parameters as is the sodium void reactivity effect in LMFBR. The steam ingress reactivity effects have been studied for 300 MWe and 1000 MWe GCFR designed at General Atomics in US., taking account of the influences of fuel burnup, fuel temperature and the presence of control rods. The calculations have been made basing on the exact perturbation theory in R-Z geometry together with JAERI-Fast 25 group constants set Version 2. In addition, to assess the uncertainties associated with the data and methods, some detailed investigations have been made on the influences of heterogeneous arrangement of fuel pins in a subassembly, and of differences in nuclear data and their processing methods to produce group constants. And also the measurements of the concerned reactivity effects at ZPR-9 have been analysed to understand our predicting accuracy. From this study, the following conclusions were drawn: (1) In the cold clean state at low fuel temperature with no control rods present, the steam ingress reactivity effect is most positive for 1000 MWe GCFR at BOL but is negative for 300 MWe GCFR. The positive reactivity effect increases as the steam density increases. (2) The presence of control rods in the core significantly reduces the large positive reactivity effect for 1000 MWe GCFR mentioned above. (author)

  13. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  14. Calculation of ex-core detector weighting functions for a sodium-cooled tru burner mockup using MCNP5

    International Nuclear Information System (INIS)

    Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A. (author)

  15. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    International Nuclear Information System (INIS)

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report

  16. Seasonal climate information preserved in West Antarctic ice core water isotopes: relationships to temperature, large-scale circulation, and sea ice

    Energy Technology Data Exchange (ETDEWEB)

    Kuettel, Marcel; Steig, Eric J.; Ding, Qinghua [University of Washington, Department of Earth and Space Sciences and Quaternary Research Center, Seattle, WA (United States); Monaghan, Andrew J. [National Center for Atmospheric Research, Boulder, CO (United States); Battisti, David S. [University of Washington, Department of Atmospheric Sciences, Seattle, WA (United States)

    2012-10-15

    As part of the United States' contribution to the International Trans-Antarctic Scientific Expedition (ITASE), a network of precisely dated and highly resolved ice cores was retrieved from West Antarctica. The ITASE dataset provides a unique record of spatial and temporal variations of stable water isotopes ({delta}{sup 18}O and {delta}D) across West Antarctica. We demonstrate that, after accounting for water vapor diffusion, seasonal information can be successfully extracted from the ITASE cores. We use meteorological reanalysis, weather station, and sea ice data to assess the role of temperature, sea ice, and the state of the large-scale atmospheric circulation in controlling seasonal average water isotope variations in West Antarctica. The strongest relationships for all variables are found in the cores on and west of the West Antarctic Ice Sheet Divide and during austral fall. During this season positive isotope anomalies in the westernmost ITASE cores are strongly related to a positive pressure anomaly over West Antarctica, low sea ice concentrations in the Ross and Amundsen Seas, and above normal temperatures. Analyses suggest that this seasonally distinct climate signal is due to the pronounced meridional oriented circulation and its linkage to enhanced sea ice variations in the adjacent Southern Ocean during fall, both of which also influence local to regional temperatures. (orig.)

  17. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  18. Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    International Nuclear Information System (INIS)

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

  19. Study on natural convection heat transfer in a vertical enclosure of double coaxial cylinder. Cooling by natural circulation of air

    International Nuclear Information System (INIS)

    To investigate a heat transfer characteristic in a vertical cavity between the pressure vessel and the cooling panel of a high-temperature engineering test reactor (HTTR), we carried out an experiment of natural convection coupled with thermal radiation in a vertical enclosure of a double coaxial cylinder. Rayleigh number based on the width of the double coaxial cylinder was set to be 5.6x105 d 8. A heat transfer coefficient of natural convection coupled with thermal radiation was obtained as function of Rayleigh number, aspect ratio of the enclosure, and the temperature of the hot and cold surface. We also carried out the numerical analysis using a heat transfer and fluid flow analytical code, which is named FLUENT/UNS. The numerical results of the temperature distribution in the apparatus showed good agreement with the experimental ones. (J.P.N.)

  20. Study on natural convection heat transfer in a vertical enclosure of double coaxial cylinder. Cooling by natural circulation of air

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Youjie [Institute on Nuclear Energy Technology, Tsinghua Univ., Beijing (China); Takeda, Tetsuaki; Inaba, Yoshitomo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2000-11-01

    To investigate a heat transfer characteristic in a vertical cavity between the pressure vessel and the cooling panel of a high-temperature engineering test reactor (HTTR), we carried out an experiment of natural convection coupled with thermal radiation in a vertical enclosure of a double coaxial cylinder. Rayleigh number based on the width of the double coaxial cylinder was set to be 5.6x10{sup 5} < Ra{sub d} < 1.04x10{sup 8}. A heat transfer coefficient of natural convection coupled with thermal radiation was obtained as function of Rayleigh number, aspect ratio of the enclosure, and the temperature of the hot and cold surface. We also carried out the numerical analysis using a heat transfer and fluid flow analytical code, which is named FLUENT/UNS. The numerical results of the temperature distribution in the apparatus showed good agreement with the experimental ones. (J.P.N.)

  1. Simulation of double cold cores of the 35°N section in the Yellow Sea with a wave-tide-circulation coupled model

    Institute of Scientific and Technical Information of China (English)

    夏长水; 乔方利; 张勐宁; 杨永增; 袁业立

    2004-01-01

    Based on the MASNUM wave-tide-circulation coupled numerical model, the temperature structure along 35°N in the Yellow Sea was simulated and compared with the observations. One of the notable features of the temperature structure along 35°N section is the double cold cores phenomena during spring and summer. The double cold cores refer to the two cold water centers located near 122°E and 125°E from the depth of 30m to bottom. The formation, maintenance and disappearance of the double cold cores are discussed. At least two reasons make the temperature in the center (near 123°E) of the section higher than that near the west and east shores in winter. One reason is that the water there is deeper than the west and east sides so its heat content is higher. The other is invasion of the warm water brought by the Yellow Sea Warm Current (YSWC) during winter.This temperature pattern of the lower layer (from 30m to bottom) is maintained through spring and summer when the upper layer (0 to 30m) is heated and strong thermocline is formed. Large zonal span of the 35°N section (about 600 km) makes the cold cores have more opportunity to survive. The double cold cores phenomena disappears in early autumn when the west cold core vanishes first with the dropping of the thermocline position.

  2. Overview of current approaches regarding the use of water to cool a molten core in the containment in ten OECD member countries

    International Nuclear Information System (INIS)

    Several countries have implemented new measures to cope with severe accidents. Of key importance is the possibility to cool the core debris in the containment and the strategy formed to accomplish this. If the core debris can be cooled pressure build-up will cease and the containment basemat will not be penetrated. The importance of this issue has been acknowledged in the development of new designs of nuclear reactors. CSNI/PWG4 task group on Containment Aspects of Severe Accident Management (CAM) started to investigate this issue in 1992. The first of a three stage approach was to investigate current accident management strategies in Member countries, identify the reasons for choosing these strategies, and consider if solutions proposed for future reactors could be applied to current reactors. A questionnaire has been sent out to the member countries. This technical report is based on the answers prepared by 10 member countries (Belgium, Finland, France, Germany, Japan, Netherlands, Spain, Sweden, Switzerland and United States), representing both BWR and PWR of different designs. In the questionnaire the member countries were asked to specify the area below the cavity, material, basement thickness, possible weaknesses, strategy chosen and the reasons for the selected strategy. The report includes three parts: current designs and strategy approaches for BWR (design of volume below reactor vessel, current approaches to cool the melted core); current designs and strategy approaches for PWR (cavity design, current approaches to cool the melted core); discussion of advantages and consequences of current strategy approaches

  3. Application of On-line Cleaning and Prefilming Technology in Refinery Circulating Cooling Water System%不停车清洗预膜技术在炼油循环冷却水系统的应用

    Institute of Scientific and Technical Information of China (English)

    2015-01-01

    Plants in refinery are various, due to the needs of production and operation, circulating cooling water system cannot be shut down for cleaning and prefilming, resulting in corrosion and fouling problems of the water cooler. The on-line cleaning and prefilming technology can realize the cleaning and prefilming of the circulating cooling water system without stopping, so that the scaling and corrosion of the circulating cooling water system can be controlled. The feasibility of the on-line cleaning and prefilming technology was investigated through using the on-line cleaning and prefilming technology in the circulating cooling water system of a refinery, and some suggestions were put forward.%炼化企业装置较多,由于生产经营的需要,循环冷却水系统可能不能停工进行清洗预膜,导致水冷器的腐蚀、结垢问题。不停车清洗预膜可以实现在不停车的情况下在线进行清洗预膜,这样就可以控制系统的结垢和腐蚀问题。某炼油厂通过不停车清洗预膜的实施效果,考察了其可行性,并提出了一些建议。

  4. Cosmic ray transport in galaxy clusters: implications for radio halos, gamma-ray signatures, and cool core heating

    Science.gov (United States)

    Enßlin, T.; Pfrommer, C.; Miniati, F.; Subramanian, K.

    2011-03-01

    We investigate the interplay of cosmic ray (CR) propagation and advection in galaxy clusters. Propagation in form of CR diffusion and streaming tends to drive the CR radial profiles towards being flat, with equal CR number density everywhere. Advection of CR by the turbulent gas motions tends to produce centrally enhanced profiles. We assume that the CR streaming velocity is of the order of the sound velocity. This is motivated by plasma physical arguments. The CR streaming is then usually larger than typical advection velocities and becomes comparable or lower than this only for periods with trans- and super-sonic cluster turbulence. As a consequence a bimodality of the CR spatial distribution results. Strongly turbulent, merging clusters should have a more centrally concentrated CR energy density profile with respect to relaxed ones with very subsonic turbulence. This translates into a bimodality of the expected diffuse radio and gamma-ray emission of clusters, since more centrally concentrated CR will find higher target densities for hadronic CR proton interactions, higher plasma wave energy densities for CR electron and proton re-acceleration, and stronger magnetic fields. Thus, the observed bimodality of cluster radio halos appears to be a natural consequence of the interplay of CR transport processes, independent of the model of radio halo formation, be it hadronic interactions of CR protons or re-acceleration of low-energy CR electrons. Energy dependence of the CR propagation should lead to spectral steepening of dying radio halos. Furthermore, we show that the interplay of CR diffusion with advection implies first order CR re-acceleration in the pressure-stratified atmospheres of galaxy clusters. Finally, we argue that CR streaming could be important in turbulent cool cores of galaxy clusters since it heats preferentially the central gas with highest cooling rate.

  5. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Chisom Shawn [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, Douglas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  6. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    International Nuclear Information System (INIS)

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  7. Investigation of primary cooling water chemistry following the partial meltdown of Pu-Be neutron source in Tehran Research Reactor Core (TRR)

    International Nuclear Information System (INIS)

    Research highlights: → Effect of Pu-Be neutron source meltdown in core on reactor water chemistry. → Water chemistry of primary cooling before, during and after of above incident was compared. → Training importance. → Management of nuclear incident and accident. - Abstract: Effect of Pu-Be neutron source meltdown in core on reactor water chemistry was main aim of this study. Leaving the neutron source in the core after reactor power exceeds a few hundred Watts was the main reason for its partial meltdown. Water chemistry of primary cooling before, during and after of above incident was compared. Activity of some radio-nuclides such as Ba-140, La-140, I-131, I-132, Te-132 and Xe-135 increased. Other radio-nuclides such as Nd-147, Xe-133, Sr-91, I-133 and I-135 are also detected which were not existed before this incident.

  8. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Short Description of the Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described, and the radiological consequences of the core modification are quantified to be tolerable

  9. Modeling of the radiation doses during dismantling of RBMK-1500 reactor emergency core cooling system large diameter pipes

    International Nuclear Information System (INIS)

    Highlights: • To propose the optimal dismantling approach, the alternatives are analyzed. • The method used of external and internal radiation assessment. • The uncertainty of collective dose and used radiation types is analyzed. • Application of the method can be extended for other types of radiation. - Abstract: Personnel radiation safety is one of the most important issues during the dismantling of nuclear installations. In this paper, results of modeling radiation doses during the dismantling of the large diameter pipes from the emergency core cooling system of RBMK-1500 reactor at Ignalina NPP are presented. The effective doses to the workers are modeled for four dismantling alternatives in order to propose the optimal dismantling approach. The impacts of the cutting technology and individual respiration protection on effective doses are analyzed. The total effective personnel doses are obtained by summing the effective personnel doses from various sources of exposure, i.e., direct radiation from radioactive equipment, internal radiation due to inhalation and ingestion of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting. The collective effective doses and their uncertainties are analyzed using VISIPLAN and MATLAB codes

  10. Ultrasonic water thickness measurement at gas-liquid interface areas in small-scale core catcher cooling experiment

    International Nuclear Information System (INIS)

    In these experiments, an ultrasonic water thickness measurement at gas-liquid interface areas, which have two types of step procedure was conducted. A the first step, its measurement was performed in the mockup condition as an ultrasonic pre-examination, and as the second step, it was performed as a pre-examination in a small-scale core catcher cooling system before a main-examination of a severe accident and PHWR safety research division at KAERI. By this method, it is possible that an ultrasonic measurement technique for determining the water layer thickness in a wavy and slug flow regime of a horizontal rectangular tube flow has been displayed in real time as A-scan mode and B-scan mode. The data of A-scanned mode were stored where the x-axis represents gap distance value, and the y-axis is only the amplitude of the interfaced area, the data of B-scanned mode can then be changed where the x-axis represents time values and the y-axis represents the value of the water thickness based on using the LabVIEW code. (author)

  11. Interactive Real-time Simulation of a Nuclear Reactor Emergency Core Cooling System on a Desktop Computer

    International Nuclear Information System (INIS)

    The simulation of the Emergency Core Cooling System for a 900 MW nuclear power plant has been developed by using object oriented programming language. It is capable of generating code that executes in real-time on a PENTIUM 100 or equivalent personal computer. Graphical user interface ECCS screens have been developed using Lab VIEW to allow interactive control of ECCS. The usual simulator functions, such as freeze, run, iterate, have been provided, and a number of malfunctions may be activated. A large pipe break near the reactor inlet header has been simulated to verify the response of the ECCS model. LOCA detection, ECC initiation, injection and recovery phased are all modeled, and give results consistent with safety analysis data for a 100% break. With stand alone ECCS simulation, the changes of flow and pressure in ECCS can be observed. The operator can study operational procedures and get used to LOCA in case of the LOCA. Practicing with malfunction, the operator will improve problem solving skills and gain a deeper comprehension of ECCS

  12. Development of Sirius facility that simulates void-reactivity feedback, and regional and core-wide stability estimation of natural circulation BWR

    International Nuclear Information System (INIS)

    The SIRIUS facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of the BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of fluid conditions, power distributions, and fuel rod thermal conductivity time constants, including the normal operating conditions of a typical natural circulation BWR. The results showed that there is a sufficiently wide stability margin under normal operating conditions, even when void-reactivity feedback is taken into account. (author)

  13. Application of the FAST code system to the static analysis of the low-void core of Gen-IV sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    The neutronic, thermal-hydraulic, and thermal-mechanic models of the low void core (CFV), a core design similar to the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) core design, represent the static core at End-of-Cycle and at nominal conditions, using the core specifications provide by the ASTRID core designers. The models are implemented in the FAST code system, a code package that establishes the coupling between 3D core neutronic, thermal-hydraulic and thermal-mechanic simulations for steady state and transient analysis. The static neutronic analysis, performed by means of the SERPENT 2 Monte-Carlo code, provides the core excess reactivity, power distributions, kinetic parameters, reactivity coefficients, and control rod reactivity worth as the main outcomes. In addition, an extensive study is carried out concerning the coolant void worth by analyzing specific sodium voiding scenarios. It can be confirmed that the ASTRID peculiar core design features a negative global coolant worth. Within the static thermal-hydraulic study, carried out by employing the TRACE code, the core assemblies division into cooling groups is performed. The respective coolant flow rates and temperatures at the core outlet are obtained. Finally, the static thermal-mechanic analysis, which was accomplished by means of the FRED code, gives general insights about the fuel temperature distribution within the fuel pins, gas gap conductance, and fission gas release. As the major outcome, the necessary static parameters of the core to proceed to the planned analyses of core transients are obtained. This document is made up of an abstract and the slides of the presentation. (author)

  14. Long-Circulating Near-Infrared Fluorescence Core-Crosslinked Polymeric Micelles: Synthesis, Characterization, and Dual Nuclear/Optical Imaging

    OpenAIRE

    Yang, Zhi; Zheng, Shiying; Harrison, William J.; Harder, John; Wen, XiaoXia; Gelovani, Juri G.; Qiao, Alex; Li, Chun

    2007-01-01

    We report the synthesis of PEG-coated, core-crosslinked polymeric micelles (CCPMs) derived from an amine-terminated amphiphilic block copolymer, poly(PEG-methacrylate)-b-poly(triethoxysilyl propylmethacrylate) (PPEGMA-b-PESPMA). The block copolymer self-assembled to form micellar nanoparticles, and a Cy-7-like near-infrared fluorescence dye was entrapped in the core bearing reactive ethoxysilane functional groups through a subsequent sol-gel process. The fluorescent signal of CCPMs on the mol...

  15. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

  16. Thermal Hydraulic Modeling of High Temperature Gas-Cooled Reactor Core in System Simulator%高温气冷堆模拟机的堆芯热工建模研究

    Institute of Scientific and Technical Information of China (English)

    孙俊; 眭喆; 张瑞鹏; 马远乐

    2012-01-01

    The nuclear power plant simulator for high temperature gas-cooled reactor (HTR) is under construction for the training and certification of the operators. The thermal hydraulic model was established in real time calculation by flow network and heat transfer network according to the core structure. In the core model, the effect of cross flow and heat transfer on the flow and temperature distributions were discussed in the normal operation and pressurized loss of forced cooling accident (PLOFC). The results indicated that the flow and heat transfer performance is not sensitive to the cross flow and heat transfer in the normal operation condition. While in the PLOFC, the cross flow and heat transfer in the reactor core is important in forming the natural circulation, removing the decay heat, as well as the transients of the flow and temperature distributions. By adding cross flow and heat transfer, the thermal states of the reactor core in the PLOFC agree well with the results by the system code. To fulfill the fidelity and scope of the HTR simulator, cross flow and heat transfer should be considered in the core model.%为完成核电厂操纵员的培训和考试,需开展适用于高温气冷堆的模拟机研究.根据球床堆芯的特点,利用流体网络与传热网络建立了可实时计算的热工水力模型,讨论了球床中氦气沿径向的流动及换热对堆芯热工水力性能的影响.结果表明,在正常运行工况下,径向流动与换热对轴向流动的影响较小,模拟结果差别不大;而在失流不失压工况下,径向流动与换热对堆芯自然对流的形成、余热导出及整个瞬态过程的影响均较为明显,考虑径向流动与换热的仿真结果与设计软件的结果符合更好.考虑到模拟机仿真范围和逼真度的要求,在高温堆模拟机堆芯建模中需加入径向流动及换热模块.

  17. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin;

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE......-like schemes in general. More importantly, we show gains of up to 4 fold over COPE-like schemes in terms of transmissions per packet in one of the investigated topologies....

  18. Effect of cooling rate on evolution of superconducting phases during decomposition and recrystallization of (Bi,Pb)-2223 core in Ag-sheathed tape

    Institute of Scientific and Technical Information of China (English)

    LI Jingyong; LI Jianguo; ZHENG Huiling; LI Chengshan; LU Yafeng; ZHOU Lian

    2006-01-01

    The reformation of (Bi,Pb)-2223 from the liquid or melt is very important for a melting process of (Bi,Pb)-2223 tape. By combination of quenching experiment with X-ray diffraction (XRD) analysis, the effect of cooling rate on the evolution of three superconducting phases in the (Bi,Pb)-2223 core of Ag-sheathed tape was investigated. The results show that (Bi,Pb)-2223 reformation from the melt seems to experience different routes during slowly cooling at different rates. One is that (Bi,Pb)-2223 phase reformed directly from the melt, and no Bi-2212 participate in this process. The other is that (Bi,Pb)-2223 is converted from the intermediate product, Bi-2212, which formed from the melt during the first cooling stage. Due to the inherent sluggish formation kinetics of (Bi,Pb)-2223 from Bi-2212, only partial (Bi,Pb)-2223 can finally be reformed with the second route.

  19. OPTICAL LINE EMISSION IN BRIGHTEST CLUSTER GALAXIES AT 0 < z < 0.6: EVIDENCE FOR A LACK OF STRONG COOL CORES 3.5 Gyr AGO?

    International Nuclear Information System (INIS)

    In recent years the number of known galaxy clusters beyond z ∼> 0.2 has increased drastically with the release of multiple catalogs containing >30,000 optically detected galaxy clusters over the range 0 0.3, hinting at an earlier epoch of strong cooling. We compare the evolution of emission-line nebulae to the X-ray-derived cool core (CC) fraction from the literature over the same redshift range and find overall agreement, with the exception that an upturn in the strong CC fraction is not observed at z > 0.3. The overall agreement between the evolution of CCs and optical line emission at low redshift suggests that emission-line surveys of galaxy clusters may provide an efficient method of indirectly probing the evolution of CCs and thus provide insights into the balance of heating and cooling processes at early cosmic times.

  20. Rapid Endovascular Catheter Core Cooling combined with cold saline as an Adjunct to Percutaneous Coronary Intervention For the Treatment of Acute Myocardial Infarction (The CHILL-MI trial)

    DEFF Research Database (Denmark)

    Erlinge, David; Götberg, Matthias; Lang, Irene;

    2014-01-01

    absolute reduction of IS/left ventricular volume of 6.2% (p = 0.15). CONCLUSIONS: Hypothermia induced by cold saline and endovascular cooling was feasible and safe, and it rapidly reduced core temperature with minor reperfusion delay. The primary end point of IS/MaR was not significantly reduced. Lower......OBJECTIVES: The aim of this study was to confirm the cardioprotective effects of hypothermia using a combination of cold saline and endovascular cooling. BACKGROUND: Hypothermia has been reported to reduce infarct size (IS) in patients with ST-segment elevation myocardial infarctions. METHODS: In a...... multicenter study, 120 patients with ST-segment elevation myocardial infarctions (<6 h) scheduled to undergo percutaneous coronary intervention were randomized to hypothermia induced by the rapid infusion of 600 to 2,000 ml cold saline and endovascular cooling or standard of care. Hypothermia was initiated...

  1. Reliability of the emergency core cooling system of the Krsko NPP in case of large-break loss of coolant accident

    International Nuclear Information System (INIS)

    Analysis of the emergency core cooling system reliability was performed for the NPP Krsko NPP with a special respect to the large-break loss-of-coolant accident. The system is divided into subsystems and then modelled using the fault tree technique. The model was analysed separately (each subsystem individually) and as a whole, considering the assumed fault criteria. In the analyses the computer codes FTAP2 and IMPORT were used. The models include human factor contributions as well. (author)

  2. Macroscopic cross sections of neutron radiation capture by Pb-208, U-238 and Tc-99 nuclides in the accelerator driven subcritical core cooled with molten Pb-208 - 286

    International Nuclear Information System (INIS)

    In the paper macroscopic cross sections for several isotopes: 208Pb, 238U, 99Tc and natural mix of lead isotopes, natPb, averaged over neutron spectra of the accelerator driven subcritical core cooled with natPb or 208Pb are given. It is shown that macro cross sections for a coolant from 208Pb are by 6.2 times smaller than those for the coolant consisted from natPb. The economy of neutrons in the core cooled with molten 208Pb can be used for reducing initial fuel load, increasing plutonium breeding and enhancing transmutation of such long lived fission products as 99Tc. The values of macro cross sections calculated for 238U and 99Tc, equal to 0.6 and 0.8 barns, respectively, are comparable with the values of the same nuclide macro cross sections for neutron spectrum of the fast reactor core cooled with sodium. Good neutron and physical features of molten 208Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. (authors)

  3. Development of small, fast reactor core designs using lead-based coolant

    International Nuclear Information System (INIS)

    A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations

  4. Evaluation of the Natural Circulation Flow Loop with Inclined Downward Heating Channel

    Energy Technology Data Exchange (ETDEWEB)

    Wi, Kyung Jin; Ha, Kwang Soon; Park, Rae Joon [KAERI, Daejeon (Korea, Republic of); Yoo, Seong Yeon [Chungnam National University, Daejeon (Korea, Republic of)

    2015-05-15

    Versatile measures have been suggested and applied to mitigate severe accidents in nuclear power plants as recently presented by Rempe et al. In general, an increase in the natural circulation mass flow rate of the coolant leads to an increase in the critical heat flux (CHF) on the hot wall, thus enhancing the thermal margin. An ex-vessel core catcher under consideration, which is one of the engineered corium cooling systems, is a passive system consisting of an inclined engineered cooling channel made of a single channel between the body of the core catcher and the inside wall of the reactor cavity. Under severe accident conditions, water is supplied from the IRWST to the engineered cooling channel. The water in the inclined channel absorbs the decay heat transferred from the corium through the carbon steel structure of the core catcher body and boils off as steam. The latter is subsequently released into the free volume of the containment above the corium spreading compartment. Water continues to flow from the IRWST to the cooling channel as a result of buoyancy-driven natural circulation. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is necessary that the water coolant be circulated at a sufficiently high rate through the inclined cooling channel for decay heat removal by downward facing boiling of the water circulated from the IRWST. KAERI performed the experimental study to evaluate the cooling performance of ex-vessel core catcher system with inclined downward facing heating surface. A scaling analysis is applied to design the test facility compared with the prototypic core catcher cooling system. The natural circulation flow experiments were performed along with the inlet subcooling, wall heat flux, and water level. The void fraction model with inclined

  5. Selected examples of natural circulation for small break loca and some severe accidents

    International Nuclear Information System (INIS)

    In all light water reactors (LWRs), natural circulation is an important passive heat removal system. The March 1979 accident at TMI-2 brought into question the capability of natural circulation cooling remove core decay heat, especially during accident situations. Because natural circulation is expected to be an essential core heat rejection mechanism during certain kinds of accidents or transients in a PWR (e.g., small break LOCAs or operational transients involving loss of pumped circulation), a thorough understanding of natural circulation processes and factors that influence the natural circulation response of the reactor system is necessary. In this paper, natural circulation and related major phenomena are discussed with examples for small break LOCA and severe accident cases, e.g., TMLB station black-out. Descriptions of three modes of natural circulation are provided: Single-phase natural circulation, two-phase natural circulation, and reflux condensation/boiling condensation. The basic phenomena associated with the three types of natural circulation being considered for severe accidents are also addressed: In-vessel natural circulation, hot leg countercurrent flow, coolant loop flows. (author)

  6. A study on the influence of boron injection tank removal on (Post) LOCA long term core cooling for Yonggwang NPP 1, 2

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jeong Jin; Song, Jong Soon [Graduate School of Chosun Univ., Kwangju (Korea, Republic of)

    2012-10-15

    The BIT (boron injection tank) of Kori NPPs 2, 3 and 4 as well as Yonggwang NPPs 1 and 2, an integral part of ECCS (emergency core cooling system), serves as its source of highly concentrated (20,000 ppm) borated water capable of 900 gallons, which is injected into a reactor coolant system via SI (safety injection) pumps. The BIT was designed to ensure emergency safety stoppage capability of the reactor in the event of MSLB (main steam line break) by preventing the sudden peak output of the core at its initial stage and continually supplying boric acid solution at a concentration of 2,400 ppm from the RWST (reactor water storage tank). Nevertheless, the present study is commenced in an attempt to know more about whether the reactor core will remain subcritical with the sufficient degree of margin during long term cooling of the core assuming that the BIT is removed, being further intended if so to be utilized for the improvement of spatial efficiency in NPPs through the rearrangement of the facility.

  7. Horizontal coring using air as the circulating fluid: Some prototype studies conducted in G Tunnel at the Nevada Test Site for the Yucca Mountain Project

    Energy Technology Data Exchange (ETDEWEB)

    Chornack, M.P. [Geological Survey, Las Vegas, NV (USA); French, C.A. [Reynolds Electrical and Engineering Co., Inc., Las Vegas, NV (USA)

    1989-12-31

    Horizontal coring using air as the circulating fluid has been conducted in the G Tunnel Underground Facility (GTUF) at the Nevada Test Site. This work is part of the prototype investigations of hydrogeology for the Yucca Mountain Project. The work is being conducted to develop methods and procedures that will be used at the Department of Energy`s Yucca Mountain Site, a candidate site for the nation`s first high-level nuclear waste repository, during the site characterization phase of the investigations. The United States Geological Survey (USGS) is conducting this prototype testing under the guidance of the Los Alamos National Laboratory (LANL) and in conjunction with Reynolds Electrical & Engineering Company (REECo), the drilling contractor. 7 refs., 8 figs., 5 tabs.

  8. COSTEAU - preheating and cooling by means of underground collectors with water circulation - case study (Perret building at Satigny, Geneva) and generalisation; COSTEAU. Prechauffage et rafraichissement par collecteurs souterrains a eau. Etude de cas (batiment Perret a Satigny, Geneve) et generalisation

    Energy Technology Data Exchange (ETDEWEB)

    Hollmuller, P.; Lachal, B.

    2003-07-01

    Since a couple of years, underground collectors with air circulation have been becoming increasingly popular as a simple means for preheating (at winter time) and cooling (at summer time) of outdoor air ahead of a ventilation system for well insulated buildings. This report considers underground collectors with water circulation used for similar purposes. They are connected to the ventilation system via an air/water heat exchanger. Starting from a case study - one-year detailed in-situ measurements and data analysis from an air-heated office building near Geneva, Switzerland - computerised simulations have been performed as a sensitivity analysis tool as well as to establish recommendations and sizing rules for planners, including cost considerations. In the case study it turned out that the water-circulated underground collector, which is installed right under the basement of this well insulated building, is in thermal contact with the basement. Its main function is to damp the daily temperature oscillation of the inlet ventilation air, bringing the expected thermal comfort improvement in the summer time. However, this underground collector is unable to collect seasonally stored heat from the ground. Hence, in the winter time the main preheating contribution arises from the series-connected heat-recovery unit from the exit air. Numerical simulations show that optimal sizing of underground collectors is essential, and that both the underground collector and the well insulated building as a physical system with thermal inertia have to be simultaneously considered in the optimization process. Optimization also has to include parasitic energy (electricity) needed by fans and pumps. As outdoor air inlet can never be flooded in the case of underground collectors with water circulation the sanitary risk encountered with air-circulated underground collectors does not exist for them. Initial investment cost for water-circulated underground collectors is higher than for a

  9. Study on in-core physical design limit zone for lead bismuth eutectic cooled long-life cycle reactor

    International Nuclear Information System (INIS)

    Reactivity variation versus core burnup is a key parameter in neutron design for long-life cycle reactor. The factors affecting reactor core loading pattern are studied from neutronic design point of view, based on the core composed of U-Pu-Zr fuel and lead bismuth eutectic coolant. The methodology for defining in-core physical design limiting zone is given, and by analyzing the effects of key parameters such as initial plutonium content and fuel rod pitch-diameter ratio, the in-core physical design limit zone is defined. Analysis results show that the methodology is appropriate and the limiting area defined in this study satisfies the core depletion and core reactivity control requirement. (authors)

  10. A Consistent Comparative Study of Advanced Sodium-cooled Fast Burner Cores loaded with Thorium and Uranium-based Metallic Fuels

    International Nuclear Information System (INIS)

    We considered uranium-based metallic fuel of TRU-U-10Zr for driver fuel and thorium was considered as blanket because thorium blanket produces less amount of TRU than uranium blanket and use of thorium blanket leads to smaller sodium void worth than the use of uranium blanket due to the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. However, these cores using thorium blanket still have a large amount of TRU production from the driver fuels because the driver fuels contain a large amount of depleted uranium which leads to the production of TRU through neutron capture. The objective of this work is to consistently compare the neutronic performances of advanced sodium cooled fast reactor cores loaded with thorium and uraniumbased metallic fuels as driver fuel for TRU burning. Our main emphasis is given on the analyses of the differences in the core performance parameters. For consistent comparison, we used the same core configuration and all the same design parameters except for the fact that depleted uranium in uraniumbased fuel is replaced with thorium. We considered the cores having no thorium blanket and the cores having thorium blanket that were designed in our previous works

  11. Nuclear Power Station Kalkar, 300 MWe Nuclear Prototype Power Station with Fast Sodium Cooled Reactor (SNR-300), Safety Report Reactor Core Mark-Ia

    International Nuclear Information System (INIS)

    The nuclear power station Kalkar is a prototype with a sodium cooled fast reactor (SNR-300) and a thermal power of 762 MW. The initial licensing procedure in 1972 was based on the so-called Mark-I core. During the following years, this core underwent some changes, for instance the thickness of the radial blanket was reduced to lower the electricity generation costs, the design of the absorber systems had been further optimized, and it became clear, that a full core with plutonium from MAGNOX-reactors could not be realized and that fuel from light-water reactors had also to be used. In this licensing document the modified reactor core Mark-Ia is described together with its assemblies and their loading procedure. The content of radioactive materials and the irradiation protection measures are discussed and those accidents are describe in an enveloping manner, from which an influence of the core modification cannot be excluded. Finally, both core versions (Mark-I and Mark-Ia) are compared with each other

  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. Study on diffusion and natural circulation of two component gases

    International Nuclear Information System (INIS)

    When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary coolant system reduces the pressure. After the reactor core and the container pressures are balanced, air is expected to ingress into the reactor core from a broken part by natural circulation and diffusion. It seems to be probable that the graphite structures is oxidized by the air. It is difficult to predict the air ingress rate, because complicated natural circulations take place in the reactor core, and diffusion paths are also complicated. In order to study the basic features of the air ingress during the early stage of the primary pipe rupture accident of the high temperature gas cooled reactor, the natural circulation of the two component gases was studied experimentally and analytically. The experiment was performed with a reversed U-shaped round pipe with one pipe heated and the other cooled. The analytical results were in good agreement with the experimental ones. (author)

  14. Influence of the adding bottom-up flow rate to the characteristic of the cooling system on TRIGA 2000 Bandung reactor core

    International Nuclear Information System (INIS)

    Heat generated from the fission reaction will heat up the cladding of the fuel element. For this reason, the fluid which used as a primary coolant in the reactor tank must have a good conductivity. This research is done to know the comparison between the performance of the natural convection cooling system and the performance of the forced convection cooling system which is done by spraying the bottom-up flow rate to the cylindrical nuclear reactor core. The result shows that the forced convection by adding spray pipe has a better performance than that of the natural convection. This case is indicated by decreasing of the maximum temperature on the top of the reactor core from 88,55°C to 47,35°C after the adding bottom-up flow rate. It can be assumed that the adding of the spraying bottom-up flow rate will give a better performance on the cooling system and will reduce the bubbles formation. (author)

  15. Specialists' meeting on gas-cooled reactor core and high temperature instrumentation, Windermere, UK, 15-17 June 1982. Summary report

    International Nuclear Information System (INIS)

    The Specialists' Meeting on ''Gas-Cooled Reactor Core and High Temperature Instrumentation'' was held at the Beech Hill Hotel, Windermere in England on June 15-17 1982. The meeting was sponsored by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors and was hosted by the Windscale Nuclear Power Development Laboratories of the UKAEA. The meeting was attended by 43 participants from Belgium, France, Federal Republic of Germany, Japan, United Kingdom of Great Britain and Northern Ireland and the United States of America. The objective of the meeting was to provide a forum, both formal and informal, for the exchange and discussion of technical information relating to instrumentation being used or under development for the measurement of core parameters, neutron flux, temperature, coolant flow etc. in gas cooled reactors. The technical part of the meeting was divided into five subject sessions: (A) Temperature Measurement (B) Neutron Detection Instrumentation (C) HTR Instrumentation - General (D) Gas Analysis and Failed Fuel Detection (E) Coolant Mass Flow and Leak Detection. A total of twenty-five papers were presented by the participants on behalf of their organizations during the meeting. A programme of the meeting and list of participants are given in appendices to this report

  16. Circulating Water Cooling Tower of Cell Transformation Ation in Sodium Metal Production%金属钠生产中电解槽循环水冷却塔改造

    Institute of Scientific and Technical Information of China (English)

    王守霞; 王广鹏

    2014-01-01

    内蒙古兰太实业股份有限公司1万t/a金属钠生产所需电解槽必须采用冷却水进行循环冷却降温,为了防止循环水管道腐蚀,确保电解槽的安全稳定运行,对循环水的水质要求较高,需要采用纯水.原有冷却塔为旧式敞开式填料冷却塔,设备老化严重,跑水量大,增加了纯水制造成本,造成水资源浪费,且敞开式的塔由于周边风沙大造成水质严重污染,泥沙经常堵塞电解槽循环水管道,也增加了工人劳动强度及清理费用.针对这一系列问题,公司建议将开式冷却塔进行改造,采用闭式冷却塔,使循环水在冷却塔换热管内循环,这样不仅避免了纯水的浪费,且大大降低了纯水的制造成本,还防止泥沙及其他杂质进入循环水系统污染水质,解决了管道堵塞的问题,降低了人员劳动强度和清理成本.%10 000 t/a metal sodium must use circulating cooling water to generate the required circulating cool for cell in Inner Mongolia Lantai Industrial Co.,Ltd..In order to prevent the circulating water pipeline corrosion,to ensure safe and stable operation of the cell,the circulating water quality required higher,which need pure water.Original cooling tower was the old open packing cooling tower,which had problem of equipment aging seriously,running large water,increasing of water production costs and resulting in waste of water resources.The big surrounding sand caused serious water pollution of the open tower.Sediment often blocked electrolyzer recycled water pipeline and also increased labor intensity and clean-up costs.For this series of questions,the company proposed to renovate the open cooling tower with closed cooling tower.The circulating water was in the cooling tower circulating tubes,which not only avoided the water waste and greatly reduced the pure water manufacturing costs,but also prevented sediment and other contaminants from entering the contaminated water circulating water system to solve

  17. An experimental study on heat transfer from a mixture of solid-fuel and liquid-steel during core disruptive accidents in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    The relocation of degraded core material through the Control Rod Guide Tubes (CRGTs) is one of essential subjects to achieve the in-vessel retention (IVR) in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The CRGT is available as the discharge path by its failure in the core region and heat-transfer from the core-material to the CRGT is one of dominant factors in its failure. In case of a core design into which a fuel subassembly with an inner duct structure (FAIDUS) is introduced, a mixture of solid-fuel and liquid-steel is supposed to remain in the core region since the FAIDUS could effectively eliminate fuel in liquid-state from the core region. Therefore, the objective of the present study is to obtain experimental knowledge for the evaluation of heat-transfer from the mixture of solid-fuel and liquid-steel to the CRGT. In the present study, an experiment was conducted using Impulse Graphite Reactor which is an experimental facility in National Nuclear Center of the Republic of Kazakhstan. In the experiment, the mixture of solid-fuel and liquid-steel was generated by a low-power nuclear heating of fuel and transferring its heat to steel, and then, data to consider the heat-transfer characteristics from the mixture of solid-fuel and liquid-steel to the CRGT were obtained. The heat-transfer characteristic was revealed by evaluating thermocouple responses observed in the experiment. Through the present study, knowledge was obtained to evaluate heat-transfer from the remaining core-materials to the CRGT. (author)

  18. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-05-15

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  19. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    International Nuclear Information System (INIS)

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  20. Changing of dominant atmospheric circulation since LGM recorded by a lake core in the central Tibetan Plateau

    Science.gov (United States)

    Zhu, L.; Wang, J.; Lu, X.; Daut, G.; Kasper, T.; Haberzettl, T.; Schwalb, A.; Maeusbacher, R.

    2013-12-01

    The mechanism of climate changes and some abrupt events on the Tibetan Plateau since LGM exists many uncertainties. Further understanding is possibly provided by a continue lake core records in the Nam Co (4718 asl, 2015 km2) on the central Tibetan Plateau. The 11m long core collected in 90m deep water area has a well age-depth distribution according to 32 14C dating data. 24-19 kaBP, higher Pediastrum suggested a shallow water condition. 19-16.5kaBP, decreased Pediastrum and Cyperaceae suggested water depth increasing and wetland reducing. Pinus, Picea and Abies were over than 30% during 24-16.5 kaBP, implying a different climate condition than it at present. 16.5-14.2 kaBP, humidity was enhanced according to Cyperaceae, Gramineae, Artemisia and Chenopodiacen. Pinus, Picea and Abies were less than 10%, suggesting climate shifted in lake area. 14.2-13.2 kaBP, Fe/Mn, Ca and Sr/Ba indicated water depth increase while total pollen concentration (TPC) and TOC (endogenesis source) reflected temperature rising. 13.2-11.5 kaBP, cold-dry climate was reflected by lake volume changing based upon Fe/Mn, Ca, Sr/Ba and Pediastrum, and the decreasing of TOC and TCP. 11.5-8.5 kaBP, a good water and heat condition was indicated by pollen assemblages and geochemistry, and the best period was within 10.2-9.3 kaBP. 8.5-5.8 kaBP, the best water-heat condition gradually weakened according to decreased TCP but stable TOC. After 5.8 kaBP, climate tended to be dry. In general, there were not only existed several climatic change events in the Nam Co lake area, but also occurred climatic type shifting since LGM.

  1. On the Fe abundance peak formation in cool-core cluster of galaxies: hints from cluster WARPJ1415.1+3612 at z=1.03

    CERN Document Server

    De Grandi, Sabrina; Nonino, Mario; Molendi, Silvano; Tozzi, Paolo; Rossetti, Mariachiara; Fritz, Alexander; Rosati, Piero

    2014-01-01

    We present a detailed study of the iron content of the core of the high redshift cluster WARPJ1415.1+3612 (z=1.03). By comparing the central Fe mass excess observed in this system, M_Fe,exc = (1.67 +\\- 0.40) x 10^9 M_sun, with those measured in local cool-core systems we infer that the bulk of the mass excess was already in place at z=1, when the age of the Universe was about half of what it is today. Our measures point to an early and intense period of star formation most likely associated with the formation of the BCG. Indeed, in the case of the power-law delay time distribution with slope -1, which best reproduces the data of WARPJ1415.1+3612, half of the supernovae explode within 0.4 Gyr, of the formation of the BCG. Finally, while for local cool-core clusters the Fe distribution is broader than the near infrared light distribution of the BCG, in WARPJ1415.1+3612 the two distributions are consistent indicating that the process responsible for broadening the Fe distribution in local systems has not yet sta...

  2. A 3D Full-Core Coupled Thermal-hydraulics/Kinetics TRACE/PARCS Model of the 2400 MWth Generation IV Gas-cooled Fast Reactor

    International Nuclear Information System (INIS)

    The present paper is related to the development and validation of a full-core coupled thermal-hydraulics (TH) and 3D kinetics TRACE/PARCS model of the large reference 2400 MWth Gas-cooled Fast Reactor (GFR) core. The GFR is an advanced fast-spectrum reactor concept currently being studied within Generation IV. This work is a preparation for the analysis of the three-dimensional core behaviour related essentially to control assembly (CA) fast movements or accidental ejections due to, for instance, the failure of a control assembly drive. The full-core model was developed using the coupled system code TRACE/PARCS, included in the FAST code system. In order to simplify the input deck preparation, specific procedures were developed and successfully used. The thermal-hydraulic and the neutronic standalone models were coupled by an external mapping scheme. Finally, coupled simulations were performed to obtain steady-state and null-transient solutions for different core configurations. The neutronics parameters, e.g. effective multiplication factor and control assembly worths, were computed and validated against static calculations performed with the deterministic system code ERANOS-2.0, good agreement being obtained in each case. (authors)

  3. Finite element based stress analysis of graphite component in high temperature gas cooled reactor core using linear and nonlinear irradiation creep models

    International Nuclear Information System (INIS)

    Highlights: • High temperature gas cooled reactor. • Finite element based stress analysis. • H-451 graphite. • Irradiation creep model. • Graphite reflector stress analysis. - Abstract: Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector

  4. Finite element based stress analysis of graphite component in high temperature gas cooled reactor core using linear and nonlinear irradiation creep models

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurindranath

    2015-10-15

    Highlights: • High temperature gas cooled reactor. • Finite element based stress analysis. • H-451 graphite. • Irradiation creep model. • Graphite reflector stress analysis. - Abstract: Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  5. HIGH-REDSHIFT X-RAY COOLING-CORE CLUSTER ASSOCIATED WITH THE LUMINOUS RADIO-LOUD QUASAR 3C 186

    International Nuclear Information System (INIS)

    We present the first results from a new, deep (200 ks) Chandra observation of the X-ray luminous galaxy cluster surrounding the powerful (L ∼ 1047 erg s-1), high-redshift (z = 1.067), compact-steep-spectrum radio-loud quasar 3C 186. The diffuse X-ray emission from the cluster has a roughly ellipsoidal shape and extends out to radii of at least ∼60 arcsec (∼500 kpc). The centroid of the diffuse X-ray emission is offset by 0.68 ± 0.''11 (∼5.5 ± 0.9 kpc) from the position of the quasar. We measure a cluster mass within the radius at which the mean enclosed density is 2500 times the critical density, r2500 = 283+18-13 kpc, of 1.02+0.21-0.14 x 1014 Msun. The gas-mass fraction within this radius is fgas = 0.129+0.015-0.016. This value is consistent with measurements at lower redshifts and implies minimal evolution in the fgas(z) relation for hot, massive clusters at 0 +0.08-0.07 Solar is consistent with the abundance observed in other massive, high-redshift clusters. The spatially resolved temperature profile for the cluster shows a drop in temperature, from kT ∼ 8 keV to kT ∼ 3 keV, in its central regions that is characteristic of cooling-core clusters. This is the first spectroscopic identification of a cooling-core cluster at z>1. We measure cooling times for the X-ray emitting gas at radii of 50 kpc and 25 kpc of 1.7 ± 0.2 x 109 years and 7.5 ± 2.6 x 108 years, as well as a nominal cooling rate (in the absence of heating) of 400 ± 190 Msun year-1 within the central 100 kpc. In principle, the cooling gas can supply enough fuel to support the growth of the supermassive black hole and to power the luminous quasar. The radiative power of the quasar exceeds by a factor of 10 the kinematic power of the central radio source, suggesting that radiative heating may be important at intermittent intervals in cluster cores.

  6. Decadal predictions of the cooling and freshening of the North Atlantic in the 1960s and the role of ocean circulation

    OpenAIRE

    Robson, Jon; Sutton, Rowan; Smith, Doug

    2014-01-01

    In the 1960s North Atlantic sea surface temperatures (SST) cooled rapidly. The magnitude of the cooling was largest in the North Atlantic subpolar gyre (SPG), and was coincident with a rapid freshening of the SPG. Here we analyze hindcasts of the 1960s North Atlantic cooling made with the UK Met Office’s decadal prediction system (DePreSys), which is initialised using observations. It is shown that DePreSys captures—with a lead time of several years—the observed cooling and freshening of the ...

  7. Finite Element Based Stress Analysis of Graphite Component in High Temperature Gas Cooled Reactor Core Using Linear and Nonlinear Irradiation Creep Models

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish; Majumdar, Saurindranath

    2015-01-01

    Irradiation creep plays a major role in the structural integrity of the graphite components in high temperature gas cooled reactors. Finite element procedures combined with a suitable irradiation creep model can be used to simulate the time-integrated structural integrity of complex shapes, such as the reactor core graphite reflector and fuel bricks. In the present work a comparative study was undertaken to understand the effect of linear and nonlinear irradiation creep on results of finite element based stress analysis. Numerical results were generated through finite element simulations of a typical graphite reflector.

  8. Using in-situ observations of atmospheric water vapor isotopes to benchmark and isotope-enabled General Circulation Models and improve ice core paleo-climate reconstruction

    Science.gov (United States)

    Steen-Larsen, Hans Christian; Sveinbjörnsdottir, Arny; Masson-Delmotte, Valerie; Werner, Martin; Risi, Camille; Yoshimura, Kei

    2016-04-01

    We have since 2010 carried out in-situ continuous water vapor isotope observations on top of the Greenland Ice Sheet (3 seasons at NEEM), in Svalbard (1 year), in Iceland (4 years), in Bermuda (4 years). The expansive dataset containing high accuracy and precision measurements of δ18O, δD, and the d-excess allow us to validate and benchmark the treatment of the atmospheric hydrological cycle's processes in General Circulation Models using simulations nudged to reanalysis products. Recent findings from both Antarctica and Greenland have documented strong interaction between the snow surface isotopes and the near surface atmospheric water vapor isotopes on diurnal to synoptic time scales. In fact, it has been shown that the snow surface isotopes take up the synoptic driven atmospheric water vapor isotopic signal in-between precipitation events, erasing the precipitation isotope signal in the surface snow. This highlights the importance of using General or Regional Climate Models, which accurately are able to simulate the atmospheric water vapor isotopic composition, to understand and interpret the ice core isotope signal. With this in mind we have used three isotope-enabled General Circulation Models (isoGSM, ECHAM5-wiso, and LMDZiso) nudged to reanalysis products. We have compared the simulations of daily mean isotope values directly with our in-situ observations. This has allowed us to characterize the variability of the isotopic composition in the models and compared it to our observations. We have specifically focused on the d-excess in order to characterize why both the mean and the variability is significantly lower than our observations. We argue that using water vapor isotopes to benchmark General Circulation Models offers an excellent tool for improving the treatment and parameterization of the atmospheric hydrological cycle. Recent studies have documented a very large inter-model dispersion in the treatment of the Arctic water cycle under a future global

  9. Melt cooling by bottom flooding: The experiment CometPC-H3. Ex-vessel core melt stabilization research

    International Nuclear Information System (INIS)

    The CometPC-H3 experiment was performed to investigate melt cooling by water addition to the bottom of the melt. The experiment was performed with a melt mass of 800 kg, 50% metal and 50% oxide, and 300 kW typical decay heat were simulated in the melt. As this was the first experiment after repair of the induction coil, attention was given to avoid overload of the induction coil and to keep the inductor voltage below critical values. Therefore, the height of the sacrificial concrete layer was reduced to 5 cm only, and the height of the porous concrete layers was also minimized to have a small distance and good coupling between heated melt and induction coil. After quite homogeneous erosion of the upper sacrificial concrete layer, passive bottom flooding started from the porous concrete after 220 s with 1.3 liter water/s. The melt was safely stopped, arrested and cooled. The porous, water filled concrete was only slightly attacked by the hot melt in the upper 25 mm of one sector of the coolant device. The peak cooling rate in the early contact phase of coolant water and melt was 4 MW/m2, and exceeded the decay heat by one order of magnitude. The cooling rate remarkably dropped, when the melt was covered by the penetrating water and a surface crust was formed. Volcanic eruptions from the melt during the solidification process were observed from 360 - 510 s and created a volcanic dome some 25 cm high, but had only minor effect on the generation of a porous structure, as the expelled melt solidified mostly with low porosity. Unfortunately, decay heat simulation in the melt was interrupted at 720 s by an incorrect safety signal, which excluded further investigation of the long term cooling processes. At that time, the melt was massively flooded by a layer of water, about 80 cm thick, and coolant water inflow was still 1 l/s. The melt had reached a stable situation: Downward erosion was stopped by the cooling process from the water filled, porous concrete layer. Top and

  10. Superconductor rotor cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Bruce B.; Sidi-Yekhlef, Ahmed; Schwall, Robert E.; Driscoll, David I.; Shoykhet, Boris A.

    2004-11-02

    A system for cooling a superconductor device includes a cryocooler located in a stationary reference frame and a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with a rotating reference frame in which the superconductor device is located. A method of cooling a superconductor device includes locating a cryocooler in a stationary reference frame, and transferring heat from a superconductor device located in a rotating reference frame to the cryocooler through a closed circulation system external to the cryocooler. The closed circulation system interfaces the stationary reference frame with the rotating reference frame.

  11. Cooling methods of station blackout scenario for LWR plants

    International Nuclear Information System (INIS)

    The objective of this study is to analyze the cooling method of station blackout scenario for both the BWR and PWR plants by RELAP5 code and to check the validity of the cooling method proposed by the utilities. In the BWR plant cooling scenario, the Reactor Core Isolation Cooling System (RCIC), which is operated with high pressure steam from the reactor, injects cooling water into the reactor to keep the core water level. The steam generated in the core is released into the suppression pool at containment vessel to condense. To restrict the containment vessel pressure rising, the ventilation from the wet-well is operated. The scenario is analyzed by RELAP5 code. In the PWR plant scenario, the primary pressure is decreased by the turbine-driven auxiliary feed water system operated with secondary side steam of the steam generators (SGs). And the core cooling is kept by the natural circulation flow at the primary loop. From the RELAP5 code analysis, it was shown that the primary system cooling was practicable by using the turbine-driven auxiliary feed water system. (author)

  12. Towards a spectroscopically accurate set of potentials for heavy hydride laser cooling candidates: Effective core potential calculations of BaH.

    Science.gov (United States)

    Moore, Keith; McLaughlin, Brendan M; Lane, Ian C

    2016-04-14

    BaH (and its isotopomers) is an attractive molecular candidate for laser cooling to ultracold temperatures and a potential precursor for the production of ultracold gases of hydrogen and deuterium. The theoretical challenge is to simulate the laser cooling cycle as reliably as possible and this paper addresses the generation of a highly accurate ab initio (2)Σ(+) potential for such studies. The performance of various basis sets within the multi-reference configuration-interaction (MRCI) approximation with the Davidson correction is tested and taken to the Complete Basis Set (CBS) limit. It is shown that the calculated molecular constants using a 46 electron effective core-potential and even-tempered augmented polarized core-valence basis sets (aug-pCVnZ-PP, n = 4 and 5) but only including three active electrons in the MRCI calculation are in excellent agreement with the available experimental values. The predicted dissociation energy De for the X(2)Σ(+) state (extrapolated to the CBS limit) is 16 895.12 cm(-1) (2.094 eV), which agrees within 0.1% of a revised experimental value of <16 910.6 cm(-1), while the calculated re is within 0.03 pm of the experimental result. PMID:27083728

  13. Suzaku X-ray Observations of the Nearest Non-Cool Core Cluster, Antlia: Dynamically Young but with Remarkably Relaxed Outskirts

    CERN Document Server

    Wong, Ka-Wah; Wik, Daniel R; Sun, Ming; Sarazin, Craig L; Fujita, Yutaka; Reiprich, Thomas H

    2016-01-01

    We present results of seven Suzaku mosaic observations (>200 ks) of the nearest non-cool core cluster, the Antlia Cluster, beyond its degree-scale virial radius (R_200) in its relaxed direction to the east. The temperature drops by a factor of three from ~2 keV near the center out to R_200, consistent with the scaled profiles of other clusters. Its pressure follows the universal profile. The density slope in its outskirts is significantly steeper than that of Virgo (a cool-core cluster with a similar temperature), but shallower than those of the massive clusters. The entropy (K) increases all the way out to R_200, consistent with the model predicted by a gravity heating-only mechanism in the outskirts. The enclosed gas mass fraction (f_gas) does not exceed the cosmic value out to 1.3 R_200. Thus, there is no evidence of significant gas clumping, electron-ion non-equipartition, or departure from the hydrostatic equilibrium (HSE) approximation that are suggested to explain the K and f_gas anomalies found in out...

  14. Proceedings of the GCNEP-IAEA course on natural circulation phenomena and passive safety systems in advanced water cooled reactors. V.2

    International Nuclear Information System (INIS)

    The current status and prospect, economics, advanced designs and applications of reactors in operation and construction, safety of advanced water cooled reactors is discussed. Papers relevant to INIS are indexed separately

  15. Hemodynamic and Thermal Responses to Head and Neck Cooling in Men and Women

    Science.gov (United States)

    Ku, Yu-Tsuan E.; Montgomery, Leslie D.; Carbo, Jorge E.; Webbon, Bruce W.

    1995-01-01

    Personal cooling systems are used to alleviate symptoms of multiple sclerosis and to prevent increased core temperature during daily activities. Configurations of these systems include passive ice vests and circulating liquid cooling garments (LCGs) in the forms of vests, cooling caps and combined head and neck cooling systems. However, little information is available oil the amount or heat that can be extracted from the body with these systems or the physiologic changes produced by routine operation of these systems. The objective of this study was to determine the operating characteristics and the physiologic change, produced by short term use of one commercially available thermal control system.

  16. Natural circulation and stratification in the various passive safety systems of the SWR 1000

    International Nuclear Information System (INIS)

    In some of the passive safety systems of Siemens' SWR 1000 boiling water reactor (i.e. the emergency condensers and containment cooling condensers), natural circulation is the main effect on both the primary and secondary sides by which optimum system efficiency is achieved. Other passive safety systems of the SWR 1000 require natural circulation on the secondary side only (condensation of steam discharged by the safety and relief valves; cooling of the Reactor Pressure Vessel (RPV) by flooding from the outside in case of core melt), while still other systems require stratification to be effective (i.e. the passive pressure pulse transmitters and steam-driven scram tanks). Complex natural circulation and stratification can take place simultaneously if fluids with different densities are enclosed in a single volume (in a core melt accident, for example, the nitrogen, steam and hydrogen in the containment). Related problems and the solutions thereto planned for the SWR 1000 are reported from the designer's viewpoint. (author)

  17. Cooling Vest

    Science.gov (United States)

    1983-01-01

    Because quadriplegics are unable to perspire below the level of spinal injury, they cannot tolerate heat stress. A cooling vest developed by Ames Research Center and Upjohn Company allows them to participate in outdoor activities. The vest is an adaptation of Ames technology for thermal control garments used to remove excess body heat of astronauts. The vest consists of a series of corrugated channels through which cooled water circulates. Its two outer layers are urethane coated nylon, and there is an inner layer which incorporates the corrugated channels. It can be worn as a backpack or affixed to a wheelchair. The unit includes a rechargeable battery, mini-pump, two quart reservoir and heat sink to cool the water.

  18. Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

  19. Natural Circulation Modeling in MTR Fuel Geometry in Research Reactors

    International Nuclear Information System (INIS)

    In many research reactors the reactor can operate under either forced or natural convection modes. Under forced convection, the primary cooling system removes the heat generated in the reactor core through a heat exchanger to the secondary cooling system, which releases this thermal energy through the cooling tower to the atmosphere. Under natural convection operating mode, the generated nuclear power imply heats up the pool water, and is ultimately dissipated through the pool surface to the containment atmosphere. Hence, the large pool provides a heat sink for the energy generated within the core. The pool surface is open to the containment atmosphere, where thermal energy exchange via evaporation and convection heat transfer can occur. In case of electrical supply failure or mechanical / electrical failure in the pumps, a Loss of Flow Accident (LOFA) will occur. In this case natural circulation mechanism may be the only mechanism that can be use to remove the decay heat from the core. Modeling and modeling validation of natural circulation is an important issue in designing and commissioning of nuclear reactors. Most of the designers are using commercial codes like RELAP5, TRAC and THYDE-W to simulate the behavior of the core under natural circulation. Validation of those cods is done by comparing between different codes and comparing the calculated results with results from scaled experimental systems. In the last two decades, CFD calculations are also used for calculations and validation of the commercial codes. In case of research reactors and in case of Loss of Flow Accident, a simple analytical model can be use for preliminary and advanced calculations of the reactors. The purpose of this paper is to present and validate a simple model for thermohydraulic calculation of research reactor in natural circulation mode

  20. Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B4C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B4C particles and the heterogeneity of the lattice cell. (author)

  1. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  2. The regulation of star formation in cool-core clusters: imprints on the stellar populations of brightest cluster galaxies

    Science.gov (United States)

    Loubser, S. I.; Babul, A.; Hoekstra, H.; Mahdavi, A.; Donahue, M.; Bildfell, C.; Voit, G. M.

    2016-02-01

    A fraction of brightest cluster galaxies (BCGs) show bright emission in the ultraviolet and the blue part of the optical spectrum, which has been interpreted as evidence of recent star formation. Most of these results are based on the analysis of broad-band photometric data. Here, we study the optical spectra of a sample of 19 BCGs hosted by X-ray luminous galaxy clusters at 0.15 BCG in Abell 1835 shows remarkable A-type stellar features indicating a relatively large population of young stars, which is extremely unusual even amongst star-forming BCGs. We constrain the mass contribution of these young components to the total stellar mass to be typically between 1 and 3 per cent, but rising to 7 per cent in Abell 1835. We find that the four of the BCGs with strong evidence for recent star formation (and only these four galaxies) are found within a projected distance of 5 kpc of their host cluster's X-ray peak, and the diffuse, X-ray gas surrounding the BCGs exhibits a ratio of the radiative cooling-to-free-fall time (tc/tff) of ≤10. These are also some of the clusters with the lowest central entropy. Our results are consistent with the predictions of the precipitation-driven star formation and active galactic nucleus feedback model, in which the radiatively cooling diffuse gas is subject to local thermal instabilities once the instability parameter tc/tff falls below ˜10, leading to the condensation and precipitation of cold gas. The number of galaxies in our sample where the host cluster satisfies all the criteria for recent and ongoing star formation is small, but their stellar populations suggest a time-scale for star formation to restart of the order of ˜200 Myr.

  3. Relocation work of temporary thermocouples for measuring the vessel cooling system in the safety demonstration test using HTTR

    International Nuclear Information System (INIS)

    High Temperature Engineering Test Reactor (HTTR), whose purpose is the establishment and upgrading of high-temperature gas-cooled reactor technology base, is the Japan's first high-temperature gas-cooled reactor that was built in the Oarai Research and Development Center of Japan Atomic Energy Agency. For the purpose of demonstrating the safety inherent in a high-temperature gas-cooled reactor, HTTR is implementing safety demonstration tests simulating the abnormal state of a reactor. In a core cooling loss test as one of the safety demonstration tests, the circulation of cooling water flowing through the vessel cooling system (VCS) is stopped. So, the temperature of the water-cooled tube panel of VCS rises, and it is necessary to monitor that the water-cooled tube panel does not exceed the maximum operation temperature during the test. Therefore, for the purpose of strengthening the monitoring VCS temperature at the time of core cooling loss test, the existing temporary thermocouples were relocated to the ring header of the outlet at the side panel of reactor vessel cooling system, and the water-cooled tube that passes through the reactor pressure vessel stabilizer. From the measurement results of temperature changes associated with the start of the reactor vessel coolant circulation pump, it was confirmed that the relocated temporary thermocouples can monitor the temperature changes of VCS. (A.O.)

  4. Chandra Observation of 3C288 - Reheating the Cool Core of a 3 keV Cluster from a Nuclear Outburst at z = 0.246

    CERN Document Server

    Lal, D V; Forman, W R; Hardcastle, M J; Jones, C; Nulsen, P E J; Evans, D A; Croston, J H; Lee, J C

    2010-01-01

    We present results from a 42 ks Chandra/ACIS-S observation of the transitional FRI/FRII radio galaxy 3C288 at z = 0.246. We detect $\\sim$3 keV gas extending to a radius of $\\sim$0.5 Mpc with a 0.5-2.0 keV luminosity of 6.6 $\\times$ 10$^{43}$ ergs s$^{-1}$, implying that 3C288 lies at the center of a poor cluster. We find multiple surface brightness discontinuities in the gas indicative of either a shock driven by the inflation of the radio lobes or a recent merger event. The temperature across the discontinuities is roughly constant with no signature of a cool core, thus disfavoring either the merger cold-front or sloshing scenarios. We argue therefore that the discontinuities are shocks due to the supersonic inflation of the radio lobes. If they are shocks, the energy of the outburst is $\\sim$10^{60} ergs, or roughly 30% of the thermal energy of the gas within the radius of the shock, assuming that the shocks are part of a front produced by a single outburst. The cooling time of the gas is $\\sim$10^8 yrs, so...

  5. The merging galaxy cluster A520 --- a broken-up cool core, a dark subcluster, and an X-ray channel

    CERN Document Server

    Wang, Qian; Giacintucci, Simona

    2016-01-01

    We present results from a deep Chandra X-ray observation of a merging galaxy cluster A520. A high-resolution gas temperature map, after the subtraction of the cluster-scale emission, reveals a long trail of dense, cool clumps --- apparently the fragments of a cool core that has been completely stripped from the infalling subcluster by ram pressure. In this scenario, we can assume that the clumps are still connected by the magnetic field lines. The observed temperature variations imply that thermal conductivity is suppressed by a factor >100 across the presumed direction of the magnetic field (as found in other clusters), and is also suppressed -along- the field lines by a factor of several. Two massive clumps in the periphery of A520, visible in the weak lensing mass map and the X-ray image, have apparently been completely stripped of gas during the merger, but then re-accreted the surrounding high-entropy gas upon exit from the cluster. An X-ray hydrostatic mass estimate for one of the clumps (that has simpl...

  6. Reference core design Mark-III of the experimental multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    The reactivity control system is one of the important items in reactor design, but it is much restricted by structural design of fuel element and pressure vessel in the experimental multi-purpose, high-temperature reactor. Preceding the first conceptual design of the reactor, therefore, the reactivity control system composed of control rod, burnable poison and reserve shutdown system in Mark-II design was re-studied, and several improvements were indicated. (1) The diameter of control rods must be as large as possible because it is impossible to increase the number of control rods. (2) The accuracy in estimation of the reactivity to be compensated with control rods is important because of the mutual interference of pair control rods with the twin configuration in a fuel element. (3) The improvement of core performance in burnup is accompanied by the reduction of design margin for control rods. (4) Increase of the reactivity to be compensated with the burnable poison leads to increase of the core reactivity recovery with burnup, and the assertion of the decrease for recovery of reactivity leads to increase of the temperature dependency of reactivity compensated with control rods. (5) Reduction of reactivity to be compensated with control rods is thus limited by cancellation of the effects in the reactivity recovery and the reactivity temperature dependency. (6) The reserve shutdown system can be designed with margin under the condition of excluding the reactivity of burnup from that to be compensated. (auth.)

  7. 高频电磁脉冲循环冷却水污垢处理与热阻监测系统%Dirt treatment by high-frequency electromagnetic pulse for circulating cooling water and thermal resistance monitoring system

    Institute of Scientific and Technical Information of China (English)

    杨子康; 熊兰; 付克勤; 苗雪飞

    2014-01-01

    The circulating cooling water systems of heat transfer equipment generates dirt easily in high temperatures.Dirt will not only reduce thermal efficiency of heat exchange equipment,but also affect the equipment safety.This paper sets up a platform of miniature water cycle based on the real structure of industrial circulating water,uses high-frequency pulse treatment method to act on calcium carbonate in circulating cooling water of manual configuration,observes the impaction of blank group and different frequencies of high frequency electromagnetic pulse groups on fouling in cooling water,and monitors the thermal resistance of cooling water by a designed thermal resistance monitoring system.The results show that the high-frequency electromagnetic pulse treatment plays a very good anti-scaling effect on circulating cooling water,because both the conductivity decline trend and pH rising trend of water becomes slow.According to the thermal resistance data of analog exchanger, thermal resistances of high frequency processing groups are less than that of blank group.Moreover,it further confirms the important role of anti-scaling for high-frequency electromagnetic pulse.%换热设备中循环冷却水系统在高温下容易产生污垢,不但会降低换热设备的换热效率,还会影响设备安全。通过模拟实际工业循环冷却水结构,搭建了微型循环冷却水实验平台,采用高频电磁脉冲对人工配置的碳酸钙循环冷却水溶液进行阻垢处理,观察空白实验组和不同频率的高频电磁脉冲处理组中水垢形成情况,并利用自行设计的热阻监测系统对管道热阻进行监测。实验结果表明:高频电磁脉冲对循环冷却水起到了很好的阻垢作用,主要表现为循环冷却水的电导率下降趋势和pH值上升趋势均变缓。根据模拟换热器的热阻监测结果显示,高频电磁脉冲处理组的热阻要小于空白实验组的热阻,并进一步证实了高频电磁脉冲的阻垢作用。

  8. Supernova Neutrino Light Curves and Spectra for Various Progenitor Stars: From Core Collapse to Proto-neutron Star Cooling

    CERN Document Server

    Nakazato, Ken'ichiro; Suzuki, Hideyuki; Totani, Tomonori; Umeda, Hideyuki; Yamada, Shoichi

    2012-01-01

    We present a new series of supernova neutrino light curves and spectra calculated by numerical simulations for a variety of progenitor stellar masses (13-50Msolar) and metallicities (Z = 0.02 and 0.004), which would be useful for a broad range of supernova neutrino studies, e.g., simulations of future neutrino burst detection by underground detectors, or theoretical predictions for the relic supernova neutrino background. To follow the evolution from the onset of collapse to 20 s after the core bounce, we combine the results of neutrino-radiation hydrodynamic simulations for the early phase and quasi-static evolutionary calculations of neutrino diffusion for the late phase, with different values of shock revival time as a parameter that should depend on the still unknown explosion mechanism. We here describe the calculation methods and basic results including the dependence on progenitor models and the shock revival time. The neutrino data are publicly available electronically.

  9. Lightweight Magnetic Cooler With a Reversible Circulator

    Science.gov (United States)

    Chen, Weibo; McCormick, John

    2011-01-01

    A design of a highly efficient and lightweight space magnetic cooler has been developed that can continuously provide remote/distributed cooling at temperatures in the range of 2 K with a heat sink at about 15 K. The innovative design uses a cryogenic circulator that enables the cooler to operate at a high cycle frequency to achieve a large cooling capacity. The ability to provide remote/distributed cooling not only allows flexible integration with a payload and spacecraft, but also reduces the mass of the magnetic shields needed. The active magnetic regenerative refrigerator (AMRR) system is shown in the figure. This design mainly consists of two identical magnetic regenerators surrounded by their superconducting magnets and a reversible circulator. Each regenerator also has a heat exchanger at its warm end to reject the magnetization heat to the heat sink, and the two regenerators share a cold-end heat exchanger to absorb heat from a cooling target. The circulator controls the flow direction, which cycles in concert with the magnetic fields, to facilitate heat transfer. Helium enters the hot end of the demagnetized column, is cooled by the refrigerant, and passes into the cold-end heat exchanger to absorb heat. The helium then enters the cold end of the magnetized column, absorbing heat from the refrigerant, and enters the hot-end heat exchanger to reject the magnetization heat. The efficient heat transfer in the AMRR allows the system to operate at a relatively short cycle period to achieve a large cooling power. The key mechanical components in the magnetic cooler are the reversible circulator and the magnetic regenerators. The circulator uses non-contacting, self-acting gas bearings and clearance seals to achieve long life and vibration- free operation. There are no valves or mechanical wear in this circulator, so the reliability is predicted to be very high. The magnetic regenerator employs a structured bed configuration. The core consists of a stack of thin

  10. Stress relaxation and creep of high-temperature gas-cooled reactor core support ceramic materials: a literature search

    International Nuclear Information System (INIS)

    Creep and stress relaxation in structural ceramics are important properties to the high-temperature design and safety analysis of the core support structure of the HTGR. The ability of the support structure to function for the lifetime of the reactor is directly related to the allowable creep strain and the ability of the structure to withstand thermal transients. The thermal-mechanical response of the core support pads to steady-state stresses and potential thermal transients depends on variables, including the ability of the ceramics to undergo some stress relaxation in relatively short times. Creep and stress relaxation phenomena in structural ceramics of interest were examined. Of the materials considered (fused silica, alumina, silicon nitride, and silicon carbide), alumina has been more extensively investigated in creep. Activation energies reported varied between 482 and 837 kJ/mole, and consequently, variations in the assigned mechanisms were noted. Nabarro-Herring creep is considered as the primary creep mechanism and no definite grain size dependence has been identified. Results for silicon nitride are in better agreement with reported activation energies. No creep data were found for fused silica or silicon carbide and no stress relaxation data were found for any of the candidate materials. While creep and stress relaxation are similar and it is theoretically possible to derive the value of one property when the other is known, no explicit demonstrated relationship exists between the two. For a given structural ceramic material, both properties must be experimentally determined to obtain the information necessary for use in high-temperature design and safety analyses

  11. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

  12. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    International Nuclear Information System (INIS)

    Highlights: ► Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. ► The two tests have been simulated by means of the Relap5 computer code. ► Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000® plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two selected tests has been defined and, in order to perform the pre-operational tests simulations, a

  13. Analysis on Cooling Capacity of Passive Core Cooling System during LOCA Scenarios%非能动堆芯冷却系统LOCA下冷却能力分析

    Institute of Scientific and Technical Information of China (English)

    游曦鸣; 邵舸; 佟立丽; 曹学武

    2016-01-01

    The analysis model for advanced pressurized water reactor was established by using mechanism analytical code .The model included reactor coolant system ,engineer‐ing safety features and related secondary side pipes system .T he typical small break loss of coolant accident and large break loss of coolant accident were selected to analyze the accident progression .The water injection capacity and cooling capacity of passive residu‐al heat removal system (PRHRS ) ,core makeup tank (CM T ) ,accumulator (ACC ) , automatic depressurization system (ADS) and in‐containment reactor water storage tank (IRWST ) included in passive core cooling system (PXS ) were focused on for LOCA with different sizes and locations .The results show that the size and location of break have an influence on the accident progression .But the peak cladding surface temperature does not exceed 1 477 K and the reactor core is flooded underwater in all the accident conditions .The PXS can effectively remove reactor core decay heat and keep the reactor in the safe shutdow n situation to prevent severe accident .%本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统(PXS)中非能动余热排出系统(PRHRS)、堆芯补水箱(CMT)、安注箱(ACC)、自动卸压系统(ADS)和安全壳内置换料水箱(IRWST)等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1477 K ,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。

  14. 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)

  15. The Application of Biocide GD-423 in the System of Circulating Cooling Water%GD-423杀菌灭藻剂在循环冷却水系统中的应用

    Institute of Scientific and Technical Information of China (English)

    文明通; 钟灵; 陈桧华; 马集锋

    2011-01-01

    GD-423是以TCMTB为主成分的新型杀菌灭藻剂,该杀菌剂对循环冷却水中的铁细菌、亚硝化菌,真菌特别有效。使用结果表明:GD-423具有杀生速度快,杀菌、灭藻及剥离粘泥效果好的特点,值得推广。%GD-423 was a kind of new biocide containing TCMTB.It has good effect on iron bacteria,nitrosomonas,fungi in circulating cooling water.The results of application show that GD-423 exhibites better biociding ability.

  16. 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.)

  17. Study on the determination of molybdate content in industrial circulating cooling water%工业循环冷却水中钼酸盐含量测定研究

    Institute of Scientific and Technical Information of China (English)

    白莹; 邵宏谦; 杨裴; 李琳

    2012-01-01

    A method for determinating the molybdate content in industrial circulating cooling water,i.e. thiocyanate spectrometry,is introduced. The molybdate is reduced to Mo5+ by adding 6.0 mL of sulfuric acid solution, 1.0 mL of ferrous ammonium sulphate solution, 10.0 mL of ammonium thiocyanate solution and 1.0 mL ascorbic acid solution, and forms orange complex with thiocyanate. The absorbency of the complex is determined at 460 run and the molybdate content is figured out by the calibration curve. This method is scientific and stable,and having strong capacity of anti-jamming. Its determination range is suitable for circulating cooling water.%介绍了一种测定工业循环冷却水中钼酸盐含量的方法——硫氰酸盐分光光度法.在待测水样中加入6.0 mL硫酸溶液,1.0 mL硫酸亚铁铵溶液,10.0 mL硫氰酸铵溶液以及1.0 mL抗坏血酸溶液,此时钼酸盐被还原成Mo5+,并与硫氰酸盐形成橙色络合物,在460m波长处测定该络合物的吸光度并通过标准曲线计算出钼酸盐含量.该方法科学、稳定,抗干扰能力强,测定范围适用于循环冷却水中钼酸盐含量的测定.

  18. Contribution to the study of the thermal and hydrodynamical properties of a two-phase natural circulation flow of normal helium (He I) for the cooling of superconducting magnets

    International Nuclear Information System (INIS)

    The method of cooling based on the thermosyphon principle is of great interest because of its simplicity, its passivity and its low cost. It is adopted to cool down to 4,5 K the superconducting magnet of the CMS particles detector of the Large Hadron Collider (LHC) experiment under construction at CERN, Geneva. This work studies heat and mass transfer characteristics of two phase He I in a natural circulation loop. The experimental set-up consists of a thermosyphon single branch loop mainly composed of a phase separator, a downward tube, and a test section. The experiments were conducted with varying several parameters such as the diameter of the test section (10 mm or 14 mm) and the applied heat flux up to the appearance of the boiling crisis. These experiments have permitted to determine the laws of evolution of the various parameters characterizing the flow (circulation mass flow rate, vapour mass flow rate, vapour quality, friction coefficient, two phase heat transfer coefficient and the critical heat flux) as a function of the applied heat flux. On the base of the obtained results, we discuss the validity of the various existing models in the literature. We show that the homogeneous model is the best model to predict the hydrodynamical properties of this type of flow in the vapour quality range 0≤x≤30%. Moreover, we propose two models for the prediction of the two phase heat transfer coefficient and the density of the critical heat flux. The first one considers that the effects of the forced convection and nucleate boiling act simultaneously and contribute to heat transfer. The second one correlates the measured critical heat flux density with the ratio altitude to diameter. (author)

  19. Circulational characteristics of a natural circulation circuit of a weakly boiling reactor large-scale model

    International Nuclear Information System (INIS)

    A large-scale model for determining circulational characteristics of a natural circulation circuit of weakly boiling (core outlet steam content below 4%) tank tpype water cooled reactors is described. The model consists of 61 elecrtroheated fuel elements 14 mm in-diameter and 3 m height. Outlet pressure can vary within 1.7-5.0 MPa inlet water subcooling is 20-90 deg C, weight outlet balance steam content from-9 to 3.2 %. Results of the experiments performed for checking the algorithms developed for thermohydraulic calculation of steady-state characteristics of the investigated circuit are given. It is concluded that for one-phase coolant estimated and experimental values for pressure head and hydraulic resistance agree well with

  20. Modular high-temperature gas-cooled reactor short term thermal response to flow and reactivity transients

    International Nuclear Information System (INIS)

    The short-term thermal response of the modular high-temperature gas-cooled reactor (MHTGR) is analyzed for a range of flow and reactivity transients. These include loss of forced circulation (LOFC) without scram, moisture ingress, spurious withdrawal of a control rod group, hypothetical large and rapid positive reactivity insertion, and a rapid core cooling event. The coupled heat transfer-neutron kinetics model is also described

  1. In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

    International Nuclear Information System (INIS)

    In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

  2. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  3. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

    2016-04-04

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  4. Code-to-code comparison for analysing the steady-state heat transfer and natural circulation in an air-cooled RCCS using GAMMA+ and Flownex

    International Nuclear Information System (INIS)

    Highlights: • The GAMMA+ and Flownex codes are used in the analyses of the air-cooled RCCS system. • Radiation heat transfer comprises the bulk of the total rate of heat transfer. • It is possible to obtain reverse flow through the RCCS standpipes. • It has been found that the results obtained with the two codes are in good agreement. • RCCS remain functional for very high blockage ratios thus supporting the safety case. - Abstract: The GAMMA+ and Flownex codes are both based on a one-dimensional flow network modelling approach and both can account for any complex network of different heat transfer phenomena occurring simultaneously. However, there are notable differences in some of the detail modelling aspects, such as the way in which the convection in the reactor cavity is represented. Despite this, it was found in the analyses of the air-cooled RCCS system that the results provided by the two codes compare very well if similar input values are used for the pressure drop coefficients, heat transfer coefficients and view factors. The results show that the radiation heat transfer comprises the bulk of the total rate of heat transfer from the RPV surface. It is also shown that it is possible to obtain a stable and sustainable steady-state operational condition where the flow is in the reverse direction through the RCCS standpipes, resulting in excessively high values for the concrete wall temperature. It is therefore crucial in the design to ensure that such a flow reversal will not occur under any circumstances. In general the good comparison between the two codes provides confidence in the ability of both to correctly solve the fundamental conservation and heat transfer relations in an integrated manner for the complete RCCS system. Provided that appropriate input values are available, these codes can therefore be used effectively to evaluate the integrated performance of the system under various operating conditions. It is shown here that the RCCS

  5. Code-to-code comparison for analysing the steady-state heat transfer and natural circulation in an air-cooled RCCS using GAMMA+ and Flownex

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, P.G., E-mail: pgr@mtechindustrial.com [School of Mechanical and Nuclear Engineering, North-West University, Private Bag X 6001, Potchefstroom (South Africa); Toit, C.G. du [School of Mechanical and Nuclear Engineering, North-West University, Private Bag X 6001, Potchefstroom (South Africa); Jun, J.S.; Noh, J.M. [Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon (Korea, Republic of)

    2015-09-15

    Highlights: • The GAMMA+ and Flownex codes are used in the analyses of the air-cooled RCCS system. • Radiation heat transfer comprises the bulk of the total rate of heat transfer. • It is possible to obtain reverse flow through the RCCS standpipes. • It has been found that the results obtained with the two codes are in good agreement. • RCCS remain functional for very high blockage ratios thus supporting the safety case. - Abstract: The GAMMA+ and Flownex codes are both based on a one-dimensional flow network modelling approach and both can account for any complex network of different heat transfer phenomena occurring simultaneously. However, there are notable differences in some of the detail modelling aspects, such as the way in which the convection in the reactor cavity is represented. Despite this, it was found in the analyses of the air-cooled RCCS system that the results provided by the two codes compare very well if similar input values are used for the pressure drop coefficients, heat transfer coefficients and view factors. The results show that the radiation heat transfer comprises the bulk of the total rate of heat transfer from the RPV surface. It is also shown that it is possible to obtain a stable and sustainable steady-state operational condition where the flow is in the reverse direction through the RCCS standpipes, resulting in excessively high values for the concrete wall temperature. It is therefore crucial in the design to ensure that such a flow reversal will not occur under any circumstances. In general the good comparison between the two codes provides confidence in the ability of both to correctly solve the fundamental conservation and heat transfer relations in an integrated manner for the complete RCCS system. Provided that appropriate input values are available, these codes can therefore be used effectively to evaluate the integrated performance of the system under various operating conditions. It is shown here that the RCCS

  6. Fetal Circulation

    Science.gov (United States)

    ... Pressure High Blood Pressure Tools & Resources Stroke More Fetal Circulation Updated:Jul 8,2016 click to enlarge The ... fetal heart. These two bypass pathways in the fetal circulation make it possible for most fetuses to survive ...

  7. Optimization of anti-fouling treatment of water of the circulation circuit (the main condenser cooling)of the Cofrentes nuclear power station

    International Nuclear Information System (INIS)

    The cooling systems is a semi-closed system with natural draft towers and a continued purge to two discharge ponds that store the water before the final discharge to the Jucar river, after having reviewed both chemically and radiologically. The total volume is 75,000 m3 and flow contribution 3,500-4,000 m3/hour, according to the seasons. It aims to minimize the content of sulfates and phosphates in the Jucar river discharge. The new treatment aims to lead the cooling water to a higher pH, decreasing the dosage of sulfuric acid and reducing organic phosphorus compounds of the components involved in the formulation of the antifouling product which is traditionally used. The content of final sulfates and phosphates in the discharge is less than that of a classic anti-fouling and consequently it obtain an environmental improvement in Jucar river discharges. The first pilot plant tests have concluded that we can, raise the average pH of 8.5 a pH average of 8.6 in the recirculating water, representing a decrease of acid consumption of 1,600 to 2,300 kg per day of 98% SO4H2 on a previous consumption of 9,000 kg per day. As to the reduction of phosphate, the new copolymer incorporates a new treatment with a higher concentration of active, it reduces the content of the phosphorus product by 29% and consequently the reduction of phosphorus in water is poured around 20% from the previous treatment. (Author)

  8. 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

  9. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    Energy Technology Data Exchange (ETDEWEB)

    Mit Basol; John F. Kielb; John F. MuHooly; Kobus Smit

    2007-05-02

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  10. Contribution to the study of the thermal and hydrodynamical properties of a two-phase natural circulation flow of normal helium (He I) for the cooling of superconducting magnets; Contribution a l'etude des proprietes thermiques et hydrodynamiques d'un ecoulement d'helium normal (He I) diphasique en circulation naturelle pour le refroidissement des aimants supraconducteurs

    Energy Technology Data Exchange (ETDEWEB)

    Benkheira, L

    2007-06-15

    The method of cooling based on the thermosyphon principle is of great interest because of its simplicity, its passivity and its low cost. It is adopted to cool down to 4,5 K the superconducting magnet of the CMS particles detector of the Large Hadron Collider (LHC) experiment under construction at CERN, Geneva. This work studies heat and mass transfer characteristics of two phase He I in a natural circulation loop. The experimental set-up consists of a thermosyphon single branch loop mainly composed of a phase separator, a downward tube, and a test section. The experiments were conducted with varying several parameters such as the diameter of the test section (10 mm or 14 mm) and the applied heat flux up to the appearance of the boiling crisis. These experiments have permitted to determine the laws of evolution of the various parameters characterizing the flow (circulation mass flow rate, vapour mass flow rate, vapour quality, friction coefficient, two phase heat transfer coefficient and the critical heat flux) as a function of the applied heat flux. On the base of the obtained results, we discuss the validity of the various existing models in the literature. We show that the homogeneous model is the best model to predict the hydrodynamical properties of this type of flow in the vapour quality range 0{<=}x{<=}30%. Moreover, we propose two models for the prediction of the two phase heat transfer coefficient and the density of the critical heat flux. The first one considers that the effects of the forced convection and nucleate boiling act simultaneously and contribute to heat transfer. The second one correlates the measured critical heat flux density with the ratio altitude to diameter. (author)

  11. Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    feedback (Doppler and displacement) VENUS-II has the following assumptions. The reactor materials behave like a homogeneous mixture with the property of an isotropic and nonviscous fluid. The reactivity change caused by a material displacement can be calculated with first-order perturbation theory. Further, the reactivity worth of spatial gradients remain constant and distort with the grid. The heat transfer from the fuel can be ignored. Although several heat transfer mechanisms can become significant, one of the greatest potential influence would appear to be a rapid molten-fuel-coolant interaction (MFCI). The non fuel core constituents are considered to be compressible, but inert, materials. The fuel vapor pressure and compression of the reactor materials are the only sources of internal pressure. Thus, such potential pressure sources such as fission gas and sodium vapor pressure are ignored. The time history of the power level can be described using point kinetics, and the spatial power-density distribution remains constant. In this work, the energy released from core disruptive accident (CDA) of sodium cooled fast reactor was investigated using CDA-ER and VENUS-II code for various reactivity insertion rates up to 100$/s, and their results were compared. The calculation results of two codes showed similar trends of energy, power and pressure from CDA. But most results of VENUS-II were found to be larger than those of CDA-ER. The released energy results calculated from VENUS-II were about 2 ∼ 3 times higher than those from CDA-ER.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-11-01

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

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

    International Nuclear Information System (INIS)

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

  14. Core meltdown assessment in the GCFR

    International Nuclear Information System (INIS)

    This paper discusses the chronological sequence of events and supporting analysis during a total loss of all coolant circulation in the GCFR with top supported core. Redundant and diverse cooling systems provided for decay heat removal reduce the probability of this postulated event below the range of plant design bases. It is nevertheless considered to investigate the potential for consequence mitigation and containment margin. Two distinct phases of the sequence are discussed: 1) the core response to a total loss of coolant circulation and 2) the capability of the PCRV to retain molten fuel debris. GCFR specific design features to prevent recriticality and fuel vaporization due to fuel slumping are under investigation. Analytical and experimental work is in progress to evaluate the feasibility of such early accident termination mechanisms. Several concepts for post accident fuel containment have been identified and appear technically feasible

  15. Emergency core cooling system simplification

    International Nuclear Information System (INIS)

    Studies and development programs at AECL over the last several years have been directed at simplification of the ECC system, with the objective of increasing reliability, reducing cost, and reducing maintenance and testing costs. This work has resulted in a substantial simplification of the ECC system for CANDU 9, including a reduction in the number of valves of over 50% relative to previous plants. This paper reviews the CANDU 9 ECC system design, and reviews the ''one-way'' rupture disc and floating ball seal development programs

  16. Application of circulating cooling water system with high cycle of concentration in treatment of high salinity water from Yellow River%高含盐黄河水高浓缩倍数运行技术及应用

    Institute of Scientific and Technical Information of China (English)

    王金华; 李本高; 傅晓萍; 李亚红

    2012-01-01

    According to the characteristics of high salinity Yellow River water and the working condition of circulating water system, a process for Yellow River water pretreatment and a high-condensive water quality stabilization technology: using mixed water (Yellow River raw water and pretreated Yellow River water) as the make-up water were developed and applied in the circulating cooling water system of a large-scale ethylene unit. The in-suit monitoring results showed that, the average corrosion rate and adhesion rate of the test tubes were 0.0186 mm/a and 5.15 mg/(cm2 · month) respectively, which were superior to the examination indexes of China Petrochemical Corporation, fully met the requirement of production facilities, and obtained a significant economic and social benefit.%针对黄河水的高含盐水质特点和循环水系统的工况条件,研究开发了黄河水预处理工艺及以黄河水混合水为循环冷却水补充水的高浓缩倍数水质稳定技术.该技术在大型乙烯装置循环冷却水系统进行工业应用,现场监测试管平均腐蚀速率为0.018 6 mm/a,平均黏附速率为5.15 mg/(cm2·月),优于中国石化的相关考核指标,完全满足生产装置对循环水的要求,具有显著的经济效益和社会效益.

  17. Two-phase natural-circulation experiments in a test facility modeled after Three Mile Island Unit-2. Final report

    International Nuclear Information System (INIS)

    A series of natural circulation experiments was conducted in a test facility that was configured after the primary and the secondary cooling systems of TMI-2. Results support the feasibility of core residual heat removal by two-phase natural circulation. Tests with noncondensable gas in the primary system indicate that two-phase natural circulation is quite tolerant of the presence of noncondensable gas. The different modes of natural circulation were discovered. Mode 1, during which only saturated steam flows in the hot leg, accomplishes the heat removal via phase changes in the vessel and in the steam generator tubes. Mode 2, during which a percolating flow exists in the hot leg, removes the heat by means of a much faster circulation in the primary loop

  18. ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

    International Nuclear Information System (INIS)

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

  19. Application of risk-informed in-service inspection approach. Pilot study on low pressure emergency core cooling system of NPP Temelin

    International Nuclear Information System (INIS)

    Core Cooling System) of Temelin NPP. Review of results, conclusions and lessons learned obtained within the Pilot study are discussed and presented. (author)

  20. Acoustical Convective Cooling Or Heating

    Science.gov (United States)

    Trinh, Eugene H.; Robey, Judith L.

    1988-01-01

    Small, efficient ultrasonic device circulates fluid. Vibrating at ultrasonic frequency, piezoelectric driver sets up vortexes transfering heat to or from object in space. Used on Earth to apply localized or concentrated cooling to individual electronic components or other small parts.

  1. Automated scoping methodology for liquid metal natural circulation small reactor

    International Nuclear Information System (INIS)

    Highlights: • Automated scoping methodology for natural circulation small modular reactor is developed. • In-house code is developed to carry out system analysis and core geometry generation during scoping. • Adjustment relations are obtained to correct the critical core geometry out of diffusion theory. • Optimized design specification is found using objective function value. • Convex hull volume is utilized to quantify the impact of different constraints on the scope range. - Abstract: A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex

  2. Examples of natural circulation in PHWR

    International Nuclear Information System (INIS)

    The main objective of this lecture is to provide deep insight into the complex natural circulation phenomena in the core of a Pressurised Heavy Water Reactor. A detailed account of natural circulation tests conducted in an Indian PHWR is given in this lecture. This will enable the participants to appreciate the importance of natural circulation in a nuclear reactor to a greater extent. (author)

  3. Lung Circulation.

    Science.gov (United States)

    Suresh, Karthik; Shimoda, Larissa A

    2016-01-01

    The circulation of the lung is unique both in volume and function. For example, it is the only organ with two circulations: the pulmonary circulation, the main function of which is gas exchange, and the bronchial circulation, a systemic vascular supply that provides oxygenated blood to the walls of the conducting airways, pulmonary arteries and veins. The pulmonary circulation accommodates the entire cardiac output, maintaining high blood flow at low intravascular arterial pressure. As compared with the systemic circulation, pulmonary arteries have thinner walls with much less vascular smooth muscle and a relative lack of basal tone. Factors controlling pulmonary blood flow include vascular structure, gravity, mechanical effects of breathing, and the influence of neural and humoral factors. Pulmonary vascular tone is also altered by hypoxia, which causes pulmonary vasoconstriction. If the hypoxic stimulus persists for a prolonged period, contraction is accompanied by remodeling of the vasculature, resulting in pulmonary hypertension. In addition, genetic and environmental factors can also confer susceptibility to development of pulmonary hypertension. Under normal conditions, the endothelium forms a tight barrier, actively regulating interstitial fluid homeostasis. Infection and inflammation compromise normal barrier homeostasis, resulting in increased permeability and edema formation. This article focuses on reviewing the basics of the lung circulation (pulmonary and bronchial), normal development and transition at birth and vasoregulation. Mechanisms contributing to pathological conditions in the pulmonary circulation, in particular when barrier function is disrupted and during development of pulmonary hypertension, will also be discussed. © 2016 American Physiological Society. Compr Physiol 6:897-943, 2016. PMID:27065170

  4. Brief Discussion on the Application of Spray Cooling Tower with Packing in the Circulation Water System in Our Company%浅谈无填料喷雾冷却塔在我公司循环水系统中的运用

    Institute of Scientific and Technical Information of China (English)

    刘宝峰; 顾收红

    2012-01-01

    This paper analyzes the application feasibility of spray cooling tower with packing in the circulation water system from the comparison of the structure,the principle and all kinds of conditions in the process of using cooling tower with packing and spray cooling tower without packing.It is hoped that cooling tower without packing be widely used in the salt making industry.%通过阐述填料冷却塔与无填料喷雾冷却塔的结构、原理以及两者在使用过程中的各种情况比较出发,深入浅出的分析了无填料喷雾冷却塔在循环水系统运用的可行性,并希望无填料塔在制盐行业中能够广泛应用。

  5. Conceptual design and safety characteristics of the natural circulation boiling water reactor HSBWR-600

    International Nuclear Information System (INIS)

    The HSBWR (Hitachi Small BWR) with a rated capacity of 600 MW electricity has been conceptually designed. The components and systems are simplified by adopting natural circulation and the passive ECCS, and eliminating steam separators. The volume of the reactor building is about 50% of that for current BWRs with the same rated capacity, and the construction period is 32-36 months until commercial operation. The major safety systems are: (1) an accumulated water injection system as an ECCS; (2) an outer pool, which stands outside of the steel primary containment vessel, as a long term cooling system after LOCAs; and (3) a steam driven reactor core isolation cooling system for high pressure water injection. The grace period is one day for core cooling and 3 days for the containment vessel heat removal. The infinite grace period for core cooling is also available as an option. LOCA analysis showed that the core will always be covered by a two-phase mixture, resulting in no core heat-up. The fundamental experiments and analyses showed sufficient capability of the outer pool for long term heat removal. (author). 12 refs, 17 figs, 3 tabs

  6. 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)

  7. The Delft desire facility for studies on (natural circulation) BWR primary system statics and dynamics

    International Nuclear Information System (INIS)

    A test facility for research on BWR core statics and dynamics was designed and built in Delft. The loop, DESIRE, consists of a BWR fuel assembly, a riser, condenser and a downcorner section. Freon-12 is used as a coolant. Presently, research on this facility is focused on investigations of the physical aspects of natural-circulation cooling and reactor kinetic stability. To this end, an artificial feedback from in-core void fraction to heating power is being established. The void fraction is determined on a sub-channel level by measuring the transmission of a collimated gamma beam

  8. Modeling strategies to compute natural circulation using CFD in a VHTR after a LOFA

    International Nuclear Information System (INIS)

    Highlights: • CFD analysis was performed of natural circulation in a block VHTR after a loss of flow accident. • Multiple strategies were investigated to perform CFD analysis to estimate strength of natural circulation. • Symmetry considerations allowed fine mesh application of CFD to reactor core. • Results extrapolated to whole core to estimate heat loss due to natural circulation. • CFD shows promise for use in nuclear reactor design and analysis. - Abstract: A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) of the U.S. Department of Energy, Office of Nuclear Energy. In the design of the prismatic VHTR, hexagonal shaped graphite blocks are drilled to allow insertion of fuel pins, made of compacted tristructural-isotropic (TRISO) fuel particles, and coolant channels for the helium coolant. One of the concerns for the reactor design is the effects of a loss of flow accident (LOFA) where the coolant circulators are lost for some reason, causing a loss of forced coolant flow through the core. In such an event, it is desired to know what happens to the (reduced) heat still being generated in the core and if it represents a problem for the fuel compacts, the graphite core or the reactor vessel (RV) walls. One of the mechanisms for the transport of heat out of the core is by the natural circulation of the coolant, which is still present. It is desired to know how much heat may be transported by natural circulation through the core and upwards to the top of the upper plenum. It is beyond current capability for a computational fluid dynamics (CFD) analysis to perform a calculation on the whole RV with a sufficiently refined mesh to examine the full potential of natural circulation in the vessel. The present paper reports the investigation of several strategies to model the flow and heat transfer in the RV. It is found that it is necessary to employ representative geometries of

  9. Thermal-hydraulic analysis for the lead-bismuth eutectic cooled reactor. System analysis by MSG-COPD code

    International Nuclear Information System (INIS)

    The feasibility study for fast breeder reactors (FBRs) including related nuclear fuel cycle systems has been started from the 1999 fiscal year by Japan Nuclear Cycle Development Institute (JNC). Phase 1 studies were finished at the end of March, 2000. Various options of FBRs plant systems was studied and concept of Lead-Bismuth Eutectic (LBE) cooled FBRs have been selected as one of these options. In the United States, the LBE cooled reactor has been examined by Generation IV. Plant dynamics analyses on 2 type of LBE-cooled reactors, forced circulation type which designed by JNC and natural circulation type which was designed by University of California, Berkeley, have been performed to understand the basic thermal-hydraulic characteristics of the reactors. As a result of the analysis on JNC forced circulation reactor, it has been clarified that hot coolant remains in the upper plenum by the thermal stratification in case of a manual trip condition. And the characteristics of pump coast down influences core exit high-temperature in case of a loss of power condition. In addition, as a result of analysis on the natural circulation reactor, the flow-redistribution effect in ductless core channels by the buoyancy force has been evaluated for a candidate duct core channels. (author)

  10. Fluid circulation and carbonate vein precipitation in the footwall of an oceanic core complex, Ocean Drilling Program Site 175, Mid-Atlantic Ridge

    Science.gov (United States)

    Schroeder, Tim; Bach, Wolfgang; Jöns, Niels; Jöns, Svenja; Monien, Patrick; Klügel, Andreas

    2015-10-01

    Carbonate veins recovered from the mafic/ultramafic footwall of an oceanic detachment fault on the Mid-Atlantic Ridge record multiple episodes of fluid movement through the detachment and secondary faults. High-temperature (˜75-175°C) calcite veins with elevated REE contents and strong positive Eu-anomalies record the mixing of up-welling hydrothermal fluids with infiltrating seawater. Carbonate precipitation is most prominent in olivine-rich troctolite, which also display a much higher degree of greenschist and sub-greenschist alteration relative to gabbro and diabase. Low-temperature calcite and aragonite veins likely precipitated from oxidizing seawater that infiltrated the detachment fault and/or within secondary faults late or post footwall denudation. Oxygen and carbon isotopes lie on a mixing line between seawater and Logatchev-like hydrothermal fluids, but precipitation temperatures are cooler than would be expected for isenthalpic mixing, suggesting conductive cooling during upward flow. There is no depth dependence of vein precipitation temperature, indicating effective cooling of the footwall via seawater infiltration through fault zones. One sample contains textural evidence of low-temperature, seawater-signature veins being cut by high-temperature, hydrothermal-signature veins. This indicates temporal variability in the fluid mixing, possibly caused by deformation-induced porosity changes or dike intrusion. The strong correlation between carbonate precipitation and olivine-rich troctolites suggests that the presence of unaltered olivine is a key requirement for carbonate precipitation from seawater and hydrothermal fluids. Our results also suggest that calcite-talc alteration of troctolites may be a more efficient CO2 trap than serpentinized peridotite.

  11. Advance in MEIC cooling studies

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yuhong [JLAB, Newport News, VA (United States); Derbenev, Ya. [JLAB, Newport News, VA (United States); Douglas, D. [JLAB, Newport News, VA (United States); Hutton, A. [JLAB, Newport News, VA (United States); Kimber, A. [JLAB, Newport News, VA (United States); Li, R. [JLAB, Newport News, VA (United States); Nissen, E. [JLAB, Newport News, VA (United States); Tennant, [JLAB, Newport News, VA (United States); Zhang, H. [JLAB, Newport News, VA (United States)

    2013-06-01

    Cooling of ion beams is essential for achieving a high luminosity for MEIC at Jefferson Lab. In this paper, we present the design concept of the electron cooling system for MEIC. In the design, two facilities are required for supporting a multi-staged cooling scheme; one is a 2 MeV DC cooler in the ion pre-booster; the other is a high electron energy (up to 55 MeV) ERL-circulator cooler in the collider ring. The simulation studies of beam dynamics in an ERL-circulator cooler are summarized and followed by a report on technology development for this cooler. We also discuss two proposed experiments for demonstrating high energy cooling with a bunched electron beam and the ERL-circulator cooler.

  12. Deep oceanic circulation in subpolar North Atlantic over the last 60 ka : a synthesis of multi-proxy approach based on Marion Dufresne cores

    Science.gov (United States)

    Kissel, Catherine; Laj, Carlo; Van Toer, Aurélie; Wandres, Camille; Michel, Elisabeth

    2015-04-01

    Different cruises on board the R. V. Marion Dufresne allowed to take cores along the paths of the main overflow waters in sub-polar North Atlantic. The cores studied for glacial period are characterized by deposition rates ranging from 8 to 24 cm/ka and those studied for the Holocene period have sedimentation rates between about 90 and 15 cm/ka. Multi-proxy approach was conducted each time with the magnetic properties as the common studied parameters, used as bottom-current tracer. These properties were coupled, depending on the cores, with oxygen and carbon isotopes of planktonic and benthic foraminifera, sortable silt, IRD counting. The rationale for the study of magnetic properties is linked to the path of the overflow waters over the sills between Greenland and Iceland and between Iceland-Faeroe and Scotland after they form in the Nordic seas. These sills are rich in magnetic particles deposited from the volcanic-rich surrounding areas and they are then more or less efficiently transported in sub-polar North Atlantic by the overflow waters depending on the intensity of the later. During the last glacial period, all the CALYPSO cores distributed from the Norwegian sea to the Bermuda Rise exhibit the same pattern of variations in magnetic concentration. The age models are based on correlation between planktonic delta18O of a core nearby Greenland and delta18O in Greenland ice (Voelker et al., 1998) and confirmed by a perfect fit between the continuous earth magnetic field intensity profile retrieved from sediments and from ice via cosmogenic isotopes. It shows that every minimum in magnetic concentration, also characterized by high IRD content, fresh surface waters, fine mean grain size in the sortable silt range, coincides with cold periods in Greenland. A synthetic "contourite drift deposit" curve has been constructed and illustrate continuously the variations in the intensity of the overflow waters during glacial time. They mimic in phase and in relative

  13. Experiment research and calculation method of natural circulation flow for AC600/1000

    International Nuclear Information System (INIS)

    Passive safety concept is extensively used in the design for next generation advanced PWR nuclear power plant. The decay heat of reactor core can be removed through natural circulation flow of coolant following an accident. This not only increases reliability of engineered safety systems and reduces core melt frequency, but also simplifies systems and increases plant economy. Nuclear Power Institute of China (NPIC) has performed preliminary experiment research and relative theoretical analysis for passive characteristics of advanced PWR nuclear power plant AC600/1000. Three tests about natural circulation flow have finished as the following: residual heat removal through SG secondary side, core makeup tank behavior and wind flow of containment. The above mentioned three mechanism tests have verified natural circulation flow concept of AC600/1000. By the end of this year NPIC will finish other two single tests in order to research the following key technology of the passive safety systems: The natural circulation characteristics of tandem system of SG secondary side loop and air flow loop for emergency residual heat removal system (ERHRS) after station blackout accident; The water flow behavior in primary coolant system contained by core makeup tank, pressurizer, accumulator and reactor pressure vessel after small break accident; Computer code development and verification. Meanwhile, NPIC will cooperate with Karlsruhe Technology Center of Germany to research natural circulation characteristics of air in the annular channel between the steel shell and the concrete shell of containment. NPIC plans to build two large integral test facilities. One of which is used to research natural circulation flow and residual heat removal through primary loop, secondary loop and air flow loop from reactor core to ultimate sink - atmosphere after station blackout accident. It is also used to research the passive safety injection features for emergency core cooling system. The second

  14. 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.)

  15. Study of natural circulation for the design of a research reactor using computational fluid dynamics and evolutionary computation techniques

    International Nuclear Information System (INIS)

    Safety is one of the most important and desirable characteristics in a nuclear plant Natural circulation cooling systems are noted for providing passive safety. These systems can be used as mechanism for removing the residual heat from the reactor, or even as the main cooling system for heated sections, such as the core. In this work, a computational fluid dynamics (CFD) code called CFX is used to simulate the process of natural circulation in a research reactor pool after its shutdown. The physical model studied is similar to the Open Pool Australian Light water reactor (OPAL), and contains the core, cooling pool, reflecting tank, circulation pipes and chimney. For best computing performance, the core region was modeled as a porous medium, where the parameters were obtained from a separately detailed CFD analysis. This work also aims to study the viability of the implementation of Differential Evolution algorithm for optimization the physical and operational parameters that, obeying the laws of similarity, lead to a test section on a reduced scale of the reactor pool.

  16. Design Construction and Operation of a Supercritical Carbon Dioxide (sCO2) Loop for Investigation of Dry Cooling and Natural Circulation Potential for Use in Advanced Small Modular Reactors Utilizing sCO2 Power Conversion Cycles.

    Energy Technology Data Exchange (ETDEWEB)

    Middleton, Bobby D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Carlson, Matthew David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    This report outlines the work completed for a Laboratory Directed Research and Development project at Sandia National Laboratories from October 2012 through September 2015. An experimental supercritical carbon dioxide (sCO 2 ) loop was designed, built, and o perated. The experimental work demonstrated that sCO 2 can be uti lized as the working fluid in an air - cooled, natural circulation configuration to transfer heat from a source to the ultimate heat sink, which is the surrounding ambient environment in most ca ses. The loop was also operated in an induction - heated, water - cooled configuration that allows for measurements of physical parameters that are difficult to isolate in the air - cooled configuration. Analysis included the development of two computational flu id dynamics models. Future work is anticipated to answer questions that were not covered in this project.

  17. Velocity Fields Measurement of Natural Circulation Flow inside a Pool Using PIV Technique

    International Nuclear Information System (INIS)

    Thermal stratification is encountered in large pool of water increasingly being used as heat sink in new generation of advanced reactors. These large pools at near atmospheric pressure provide a heat sink for heat removal from the reactor or steam generator, and the containment by natural circulation as well as a source of water for core cooling. For examples, the PAFS (passive auxiliary feedwater system) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus), which is intended to completely replace the conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism. In a pool, the heat transfer from the PCHX (passive condensation heat exchanger) contributed to increase the pool temperature up to the saturation condition and induce the natural circulation flow of the PCCT (passive condensate cooling tank) pool water. When a heat rod is placed horizontally in a pool of water, the fluid adjacent to the heat rod gets heated up. In the process, its density reduces and by virtue of the buoyancy force, the fluid in this region moves up. After reaching the top free surface, the heated water moves towards the other side wall of the pool along the free surface. Since this heated water is cooling, it goes downward along the wall at the other side wall. Above heater rod, a natural circulation flow is formed. However, there is no flow below heater rod until pool water temperature increases to saturation temperature. In this study, velocity measurement was conducted to reveal a natural circulation flow structure in a small pool using PIV (particle image velocimetry) measurement technique

  18. Prediction of boiling-induced natural circulation flow in an inclined channel with non-uniform flow area

    International Nuclear Information System (INIS)

    The boiling-induced natural circulation flow in the engineered cooling channel is modelled and solved by considering the conservation of mass, momentum and energy in the two-phase mixture, along with the two-phase friction drop and void fraction. The model has been applied to estimate the induced mass flow rates through a uniform and non-uniform annular gap between the reactor vessel and insulation under the IVR-ERVC conditions, and also the engineered corium cooling system of an ex-vessel core catcher during a severe accident for various system parameters including the channel gap size, inlet diameter, inlet subcooling, and wall heat flux. (author)

  19. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  20. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    International Nuclear Information System (INIS)

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm2, 10000C cladding temperature, and (2) 40 h at 40 W/cm2, 12000C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 13700C

  1. Sample of EDF-R&D 2009-2012 core design studies on: Heterogeneous sodium-cooled fast reactors with low sodium void effect

    International Nuclear Information System (INIS)

    • The use of axial or radial heterogeneities allows to reach negative global sodium void worth: → 600 MWe radially heterogeneous cores can be designed (no annual shape, reasonable core size). But the constraint on the fissile radius might be the reason why current performances are not a breakthrough... → Other interesting configuration exists (standard 2 diabolos, etc.). • Complexity versus intrinsic safety: where should the line be drawn? → On 1200 MWe cores, it was shown that very good features can be reached. But very complex designs are needed. • Core design studies are still going on... → The precision of global optimization method will be improved and new safety indicators added. → Optimization studies will be led on the CEA CFV design. → (Little) time should be spent on alternative design options

  2. Transient following partial loss of feed water for thorium based natural circulation reactor

    International Nuclear Information System (INIS)

    The proposed Advanced Heavy Water Reactor (AHWR) is a 920 MWth vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all allowed power levels with no primary coolant pumps. Apart from this passive design feature passive safety systems in AHWR include isolation condenser (IC) system for decay heat removal in case of unavailability of main steam condenser, emergency core cooling (includes both high pressure and low pressure ECC) system, Passive containment cooling system, Passive containment isolation and automatic depressurization system. Further, reactor core has negative void coefficient of reactivity at all power level which enhances the safety of the reactor. The Primary Heat Transport System of the reactor consists of reactor core, core inlet and outlet core bottom extensions, inlet feeders, tailpipes, steam drums, downcomer and inlet header. One BFP trip transient without standby pump initiation has been analysed. This partial loss of feed scenario leads to reactor trip and subsequent decay heat removal takes place through isolation condenser path. All other thermal hydraulic parameters remain within safety limits

  3. Preliminary Analysis on Reactivity Insertion Transient of Natural Circulation Reactor

    International Nuclear Information System (INIS)

    When a malfunction of the reactor control system occurs, there's a chance that the positive reactivity is inserted into the core, resulting in the increase of the core power. With the combination of the failure of the related safety features, this may raise the temperature of the core material beyond the design limit to break its integrity. For the fast nuclear reactors like FFTF and CRBRP, the overpower trip is initiated when the power reaches 115% of rated value to keep the fuel from melting. In this study, the system response to the reactivity insertion transient on a liquid metal cooled natural circulation reactor is analyzed utilizing an in-house code based on a momentum integral model. Utilizing an in-house system analysis code, a set of numerical simulation is carried out on the reactivity ramp insertion transients which showed that the role of reactivity feedback is significant in mitigating the time to failure, and the evolution of the natural circulation mass flow is rather slow to generate meaningful feedback effect. It is also observed that in terms of the peak cladding temperature, smaller reactivity insertion transient generated more severe outcome owing to increased accumulation of the thermal energy within fuel pins. Thus, it may require an extra attention to carefully monitor and capture the mild transients to avoid potential drastic results

  4. Sample of EDF-R&D 2009-2012 core studies on heterogeneous sodium-cooled fast reactors with low sodium void effect

    International Nuclear Information System (INIS)

    The unprotected loss of primary or/and secondary pumps transient is often used as a reference initiator to evaluate the intrinsic behaviour of SFR cores. Traditional designs show a major positive sodium reactivity feedback during this transient and many studies were led in the past to get rid of that effect. In 2009-2012, consistently with researches led at CEA and AREVA, cores with low sodium void effect were studied and optimized at EDF-R&D. The use of a sodium plenum with a boron carbide plate, a reduction of the fissile height, or a step-wise modulation of the fissile height can be used to reduce the sodium void worth. To go further, neutronic leakage can be enhanced using axial or radial heterogeneities. Axial heterogeneity is the basis of the CFV CEA concept. An analysis of this design using an EDF optimization methodology led to a better understanding and allowed to define core variants. As far as radial heterogeneities are concerned, various configurations featuring annular shapes, fertile sub-assembly rings and modulations of the fissile height are presented in this paper. One of these concepts is applied on a 600 MWe scale and compared to a CFV-like core. This paper aims at giving an overview of EDF-R&D core design studies on the time period 2009-2012, and concludes on perspectives for future work. (author)

  5. Investigation of interaction between heat transport systems during the natural circulation decay heat removal in FBRs. Influence of decay heat removal system type and the secondary heat transport system

    International Nuclear Information System (INIS)

    Steady state sodium experiments were performed to investigate interactions between the heat transport systems, i.e., the primary system, the secondary system, and the decay heat removal system, during the natural circulation decay heat removal in FBRs. The test rig was used for the experiments. The core model has seven subassemblies; the center assembly simulates pin bundle geometry of a core fuel subassembly in a large scale FBR and consists of 37 pins, six outer subassemblies consists of 7 pins. As the decay heat removal system, Direct Reactor Auxiliary Cooling System (DRACS) and Primary Reactor Auxiliary Cooling System (PRACS) can be selected. Experiments were carried out under natural circulation conditions in the primary loop and force convection conditions in the decay heat removal system. In cases using DRACS, natural circulation flow rate in the primary loop was smaller by 20% than that in cases using PRACS due to the low temperature in the upper plenum and also in the upper non-heated section of the core. When natural circulation was allowed in the secondary heat transport system, the natural circulation flow rate in the primary system increased in spite of the operation of DRACS. In cases using DRACS, inter-subassembly flow redistribution occurred; the center subassembly had larger flow rate than those in outer subassemblies due to the low natural circulation head in the outer subassemblies which were cooled by the inter-wrapper flow (IWF). The highest temperature in the core was reduced by IWF via not only the direct cooling effect but also the inter-subassembly flow redistribution. (J.P.N.)

  6. Second sector cool down

    CERN Multimedia

    2007-01-01

    At the beginning of July, cool-down is starting in the second LHC sector, sector 4-5. The cool down of sector 4-5 may occasionally generate mist at Point 4, like that produced last January (photo) during the cool-down of sector 7-8.Things are getting colder in the LHC. Sector 7-8 has been kept at 1.9 K for three weeks with excellent stability (see Bulletin No. 16-17 of 16 April 2007). The electrical tests in this sector have got opt to a successful start. At the beginning of July the cryogenic teams started to cool a second sector, sector 4-5. At Point 4 in Echenevex, where one of the LHC’s cryogenic plants is located, preparations for the first phase of the cool-down are underway. During this phase, the sector will first be cooled to 80 K (-193°C), the temperature of liquid nitrogen. As for the first sector, 1200 tonnes of liquid nitrogen will be used for the cool-down. In fact, the nitrogen circulates only at the surface in the ...

  7. MEIC electron cooling program

    International Nuclear Information System (INIS)

    Cooling of proton and ion beams is essential for achieving high luminosities (up to above 1034 cm-2s-1) for MEIC, a Medium energy Electron-Ion Collider envisioned at JLab [1] for advanced nuclear science research. In the present conceptual design, we utilize the conventional election cooling method and adopted a multi-staged cooling scheme for reduction of and maintaining low beam emittances [2,3,4]. Two electron cooling facilities are required to support the scheme: one is a low energy (up to 2 MeV) DC cooler installed in the MEIC ion pre-booster (with the proton kinetic energy up to 3 GeV); the other is a high electron energy (up to 55 MeV) cooler in the collider ring (with the proton kinetic energy from 25 to 100 GeV). The high energy cooler, which is based on the ERL technology and a circulator ring, utilizes a bunched electron beam to cool bunched proton or ion beams. To complete the MEIC cooling concept and a technical design of the ERL cooler as well as to develop supporting technologies, an R&D program has been initiated at Jefferson Lab and significant progresses have been made since then. In this study, we present a brief description of the cooler design and a summary of the progress in this cooling R&D

  8. Prevention and investigations of core degradation in case of beyond design accidents of the 2400 MWTH gas-cooled fast reactor

    International Nuclear Information System (INIS)

    The present paper deals with studies carried out to assess the ability of the core of the Gas Fast Reactor (GFR) to withstand beyond design accidents. The work presented here is aimed at simulating the behaviour of this core by using analytical models whose input parameters are calculated with the CATHARE2 code. Among possible severe accident initiators, the Unprotected Loss Of Coolant Accident (ULOCA of 3 Inches diameter) is investigated in detail in the paper with CATHARE2. Additionally, a simplified pessimistic assessment of the effect of a postulated power excursion that could result from the failure of prevention provisions is presented. (author)

  9. Boiling induced mixed convection in cooling loops

    International Nuclear Information System (INIS)

    This article describes the SUCO program performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. In case of a core melt accident, the sump cooling concept realises a decay heat removal system that is based on passive safety features within the containment. The article gives, first, results of the experiments in the 1:20 linearly scaled SUCOS-2D test facility. The experimental results are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. Second, the real height SUCOT test facility with a volume and power scale of 1:356 that is aimed at investigating the mixed single-phase and two-phase natural circulation flow in the reactor sump, together with first measurement results, are discussed. Finally, a numerical approach to model the subcooled nucleate boiling phenomena in the test facility SUCOT is presented. Physical models describing interfacial mass, momentum and-heat transfer are developed and implemented in the commercial software package CFX4.1. The models are validated for an isothermal air-water bubbly flow experiment and a subcooled boiling experiment in vertical annular water flow. (author)

  10. Stochastic Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Blaskiewicz, M.

    2011-01-01

    Stochastic Cooling was invented by Simon van der Meer and was demonstrated at the CERN ISR and ICE (Initial Cooling Experiment). Operational systems were developed at Fermilab and CERN. A complete theory of cooling of unbunched beams was developed, and was applied at CERN and Fermilab. Several new and existing rings employ coasting beam cooling. Bunched beam cooling was demonstrated in ICE and has been observed in several rings designed for coasting beam cooling. High energy bunched beams have proven more difficult. Signal suppression was achieved in the Tevatron, though operational cooling was not pursued at Fermilab. Longitudinal cooling was achieved in the RHIC collider. More recently a vertical cooling system in RHIC cooled both transverse dimensions via betatron coupling.

  11. Development of local heat transfer models for the safety assessment of prismatic modular high temperature gas-cooled reactor cores - HTR2008-58295

    International Nuclear Information System (INIS)

    This paper presents a model developed for determining fuel particle and fuel block temperatures of a prismatic core modular reactor during both normal operation and under fault conditions. The model is based on multi-scale modeling techniques and has been qualified by comparison with finite element simulations for both steady state and transient conditions. Further, a model for determining the effective conductivity of the block fuel elements - important for heat removal in loss of flow conditions - is presented and, again, qualified by comparison with finite element simulations. A numerical model for predicting conduction heat transfer both within and between block fuel elements has been developed which, when coupled with the above multi-scale model, allows simulations of whole cores to be carried out whilst retaining the ability to predict the temperatures of individual coolant channels and individual coated particles in the fuel if required. This ability to resolve heat transfer on length scales ranging from a few meters down to a few microns within the same model is very powerful and allows a complete assessment of the fuel and structural temperatures within a core to be made. More significantly, this level of resolution facilitates interactive coupling with neutronics models to enable the strong temperature/reactivity feedbacks, inherent in such cores, to be resolved correctly. (authors)

  12. Core Overheating Event Probability Analysis due Lack of Passive Cooling in a Typical MTR Pool Type Research Reactor by FTA Study

    International Nuclear Information System (INIS)

    Through the presentation, the accidental top event of core overheating after normal shutdown is analyzed by a FTA diagram down to primary basic events, for which the probability of occurrence is provided. Finally, two main outputs of the analysis will be presented

  13. 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

  14. Enhancement of the southward return flow of the Atlantic Meridional Overturning Circulation by data assimilation and its influence in an assimilative ocean simulation forced by CORE-II atmospheric forcing

    Science.gov (United States)

    Fujii, Yosuke; Tsujino, Hiroyuki; Toyoda, Takahiro; Nakano, Hideyuki

    2015-08-01

    This paper examines the difference in the Atlantic Meridional Overturning Circulation (AMOC) mean state between free and assimilative simulations of a common ocean model using a common interannual atmospheric forcing. In the assimilative simulation, the reproduction of cold cores in the Nordic Seas, which is absent in the free simulation, enhances the overflow to the North Atlantic and improves AMOC with enhanced transport of the deeper part of the southward return flow. This improvement also induces an enhanced supply of North Atlantic Deep Water (NADW) and causes better representation of the Atlantic deep layer despite the fact that correction by the data assimilation is applied only to temperature and salinity above a depth of 1750 m. It also affects Circumpolar Deep Water in the Southern Ocean. Although the earliest influence of the improvement propagated by coastal waves reaches the Southern Ocean in 10-15 years, substantial influence associated with the arrival of the renewed NADW propagates across the Atlantic Basin in several decades. Although the result demonstrates that data assimilation is able to improve the deep ocean state even if there is no data there, it also indicates that long-term integration is required to reproduce variability in the deep ocean originating from variations in the upper ocean. This study thus provides insights on the reliability of AMOC and the ocean state in the Atlantic deep layer reproduced by data assimilation systems.

  15. Cooling rates and the depth of detachment faulting at oceanic core complexes: Evidence from zircon Pb/U and (U-Th)/He ages

    Science.gov (United States)

    Grimes, Craig B.; Cheadle, Michael J.; John, Barbara E.; Reiners, P.W.; Wooden, J.L.

    2011-01-01

    Oceanic detachment faulting represents a distinct mode of seafloor spreading at slow spreading mid-ocean ridges, but many questions persist about the thermal evolution and depth of faulting. We present new Pb/U and (U-Th)/He zircon ages and combine them with magnetic anomaly ages to define the cooling histories of gabbroic crust exposed by oceanic detachment faults at three sites along the Mid-Atlantic Ridge (Ocean Drilling Program (ODP) holes 1270D and 1275D near the 15??20???N Transform, and Atlantis Massif at 30??N). Closure temperatures for the Pb/U (???800??C-850??C) and (U-Th)/He (???210??C) isotopic systems in zircon bracket acquisition of magnetic remanence, collectively providing a temperature-time history during faulting. Results indicate cooling to ???200??C in 0.3-0.5 Myr after zircon crystallization, recording time-averaged cooling rates of ???1000??C- 2000??C/Myr. Assuming the footwalls were denuded along single continuous faults, differences in Pb/U and (U-Th)/He zircon ages together with independently determined slip rates allow the distance between the ???850??C and ???200??C isotherms along the fault plane to be estimated. Calculated distances are 8.4 ?? 4.2 km and 5.0 2.1 km from holes 1275D and 1270D and 8.4 ?? 1.4 km at Atlantis Massif. Estimating an initial subsurface fault dip of 50 and a depth of 1.5 km to the 200??C isotherm leads to the prediction that the ???850??C isotherm lies ???5-7 km below seafloor at the time of faulting. These depth estimates for active fault systems are consistent with depths of microseismicity observed beneath the hypothesized detachment fault at the TAG hydrothermal field and high-temperature fault rocks recovered from many oceanic detachment faults. Copyright 2011 by the American Geophysical Union.

  16. Cooled water rod (loca conditions)

    International Nuclear Information System (INIS)

    A process is described for providing a radiation heat sink for fuel bundles having a large water moderator tube in the event of a loss of coolant accident the fuel bundles having an upper tie plate, a lower tie plate, a channel surrounding and connecting the tie plate, a plurality of fuel rods supported between the tie plates and within the channels in side by side upstanding relation; a large water moderator tube having at least twice the diameter of the fuel rods. The process consists of: spraying core cooling spray in an evenly divided flow over the upper tie plate; collecting core cooling spray at an uper end of the large water moderator tube; and distributing the core cooling spray circumferentially along the inner surfaces of the large water moderator tube in a downward flow separating the flow of the core cooling spray from the flow of steam resulting from the flashing of water to steam within the moderator tube

  17. Quantities of I-131 and Cs-137 in accumulated water in the basements of reactor buildings in process of core cooling at Fukushima Daiichi nuclear power plants accident and its influence on late phase source terms

    International Nuclear Information System (INIS)

    During the process of core cooling at Fukushima Daiichi nuclear power plants accident, large amount of contaminated water was accumulated in the basements of the reactor buildings at Units 1-4. The present study estimated the quantities of I-131 and Cs-137 in the water as of late March based on the press-opened data. The estimated ratios of I-131 and Cs-137 quantities to the core inventories are 0.51%, 0.85% at Unit 1, 74%, 38% at Unit 2 and 26%, 18% at Unit 3, respectively. According to the Henry's law, certain fraction of iodine in water could be released to atmosphere due to gas-liquid partition and contribute to increase in the release to environment. A lot of evaluations for I-131 release have been performed so far by the MELCOR calculation or the SPEEDI reverse estimation. The SPEEDI reverse predicted significant release until 26 March, while no prediction in MELCOR after 17 March. The present study showed that iodine release from accumulated water may explain the release between 17 and 26 March. This strongly suggests a need for improvement of current MELCOR approach which treats the release only from containment breaks for several days after the core melt. (author)

  18. Stacking with stochastic cooling

    International Nuclear Information System (INIS)

    Accumulation of large stacks of antiprotons or ions with the aid of stochastic cooling is more delicate than cooling a constant intensity beam. Basically the difficulty stems from the fact that the optimized gain and the cooling rate are inversely proportional to the number of particles 'seen' by the cooling system. Therefore, to maintain fast stacking, the newly injected batch has to be strongly 'protected' from the Schottky noise of the stack. Vice versa the stack has to be efficiently 'shielded' against the high gain cooling system for the injected beam. In the antiproton accumulators with stacking ratios up to 105 the problem is solved by radial separation of the injection and the stack orbits in a region of large dispersion. An array of several tapered cooling systems with a matched gain profile provides a continuous particle flux towards the high-density stack core. Shielding of the different systems from each other is obtained both through the spatial separation and via the revolution frequencies (filters). In the 'old AA', where the antiproton collection and stacking was done in one single ring, the injected beam was further shielded during cooling by means of a movable shutter. The complexity of these systems is very high. For more modest stacking ratios, one might use azimuthal rather than radial separation of stack and injected beam. Schematically half of the circumference would be used to accept and cool new beam and the remainder to house the stack. Fast gating is then required between the high gain cooling of the injected beam and the low gain stack cooling. RF-gymnastics are used to merge the pre-cooled batch with the stack, to re-create free space for the next injection, and to capture the new batch. This scheme is less demanding for the storage ring lattice, but at the expense of some reduction in stacking rate. The talk reviews the 'radial' separation schemes and also gives some considerations to the 'azimuthal' schemes

  19. 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)

  20. The Effect of Medium Leakage on the Microbial Diversity of Circulating Cooling Water%介质泄漏对循环冷却水中微生物多样性的影响

    Institute of Scientific and Technical Information of China (English)

    董文文; 刘芳; 仲慧赟; 卢宪辉; 陆津津

    2013-01-01

    以循环冷却水作为接种水对生物粘泥进行培养,向循环冷却水中加入柴油以模拟炼油厂中的介质泄漏现象,对介质泄漏影响下生物粘泥中的微生物进行微观分析,利用扫描电子显微镜(SEM)观察生物粘泥内部的空间结构、紧密度等,利用聚合酶链式反应-变性梯度凝胶电泳(PCR-DGGE)技术,分析不同生物粘泥中的内部优势菌种、微生物多样性及相似性.SEM分析表明,与未投油的生物粘泥比较,投加0.3g·L-1柴油时的生物粘泥内部结构复杂、紧密度好,而投加0.9g·L-1柴油时的生物粘泥内部结构简单、紧密度差.PCR-DGGE分析表明,与投加0.9g·L-1柴油的生物污泥相比,投加0.3g·L-1柴油的生物粘泥的细菌数量和种类更多,微生物多样性更大,优势菌种更多.%The slime was cultured with the circulating cooling water as the inoculated water,adding diesel to analog medium leakage of refineries. Microorganisms in the slime under the influence of medium leakage were analyzed,and the structure,density,thickness and compactness inside the slime were observed with scanning e-lectron microscope(SEM). The dominant species,microbial diversity and similarity of different slimes were analyzed by using polymerase chain reaction-denaturing gradient gel electrophoresis(PCR-DGGE). The SEM results showed that campared with the slime without adding diesel, the slime adding 0. 3 g · L-1 of diesel had complex structure and higher compactness,but the slime adding 0. 9 g · L-1 of diesel had simple structure and lower compactness. The PCR-DGGE result showed that,quantity and types of bacteria,the microbial diversity and the dominant species amount of the slime adding 0. 3 g · L-1 of diesel were all superior to those of that adding 0. 9 g · L-1 of diesel.