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Sample records for core cooling system

  1. Emergency core cooling system

    International Nuclear Information System (INIS)

    Ando, Masaki.

    1987-01-01

    Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

  2. Core cooling systems

    International Nuclear Information System (INIS)

    Hoeppner, G.

    1980-01-01

    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)

  3. Emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Nobuaki.

    1993-01-01

    A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

  4. Emergency core cooling system

    International Nuclear Information System (INIS)

    Kato, Ken.

    1989-01-01

    In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

  5. Emergency core cooling system

    International Nuclear Information System (INIS)

    Sato, Akira; Kobayashi, Masahide.

    1983-01-01

    Purpose: To enable a stable operation of an emergency core cooling system by preventing the system from the automatic stopping at an abnormally high level of the reactor water during its operation. Constitution: A pump flow rate signal and a reactor water level signal are used and, when the reactor water level is increased to a predetermined level, the pump flow rate is controlled by the reactor water level signal instead of the flow rate signal. Specifically, when the reactor water level is gradually increased by the water injection from the pump and exceeds a setting signal for the water level, the water level deviation signal acts as a demand signal for the decrease in the flow rate of the pump and the output signal from the water level controller is also decreased depending on the control constant. At a certain point, the output signal from the water level controller becomes smaller than the output signal from the flow rate controller. Thus, the output signal from the water level controller is outputted as the output signal for the lower level preference device. In this way, the reactor water level and the pump flow rate can be controlled within a range not exceeding the predetermined pump flow rate. (Horiuchi, T.)

  6. Emergency core cooling system

    International Nuclear Information System (INIS)

    Arai, Kenji; Oikawa, Hirohide.

    1990-01-01

    The device according to this invention can ensure cooling water required for emerency core cooling upon emergence such as abnormally, for example, loss of coolant accident, without using dynamic equipments such as a centrifugal pump or large-scaled tank. The device comprises a pressure accumulation tank containing a high pressure nitrogen gas and cooling water inside, a condensate storage tank, a pressure suppression pool and a jet stream pump. In this device there are disposed a pipeline for guiding cooling water in the pressure accumulation tank as a jetting water to a jetting stream pump, a pipeline for guiding cooling water stored in the condensate storage tank and the pressure suppression pool as pumped water to the jetting pump and, further, a pipeline for guiding the discharged water from the jet stream pump which is a mixed stream of pumped water and jetting water into the reactor pressure vessel. In this constitution, a sufficient amount of water ranging from relatively high pressure to low pressure can be supplied into the reactor pressure vessel, without increasing the size of the pressure accumulation tank. (I.S.)

  7. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

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

  8. Emergency core cooling systems

    International Nuclear Information System (INIS)

    Kubokoya, Takashi; Okataku, Yasukuni.

    1984-01-01

    Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

  9. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

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

    1971-01-01

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

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

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

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

  11. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

    This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

  12. Reactor-core isolation cooling system with dedicated generator

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  15. Emergency core cooling system in BWR type reactors

    International Nuclear Information System (INIS)

    Takizawa, Yoji

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  17. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... comments were received. A companion guide, DG-1277, ``Initial Test Program of Emergency Core Cooling... NUCLEAR REGULATORY COMMISSION [NRC-2011-0129] Preoperational Testing of Emergency Core Cooling... (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

  18. Unravelling the core microbiome of biofilms in cooling tower systems.

    Science.gov (United States)

    Di Gregorio, L; Tandoi, V; Congestri, R; Rossetti, S; Di Pippo, F

    2017-11-01

    In this study, next generation sequencing and catalyzed reporter deposition fluorescence in situ hybridization, combined with confocal microscopy, were used to provide insights into the biodiversity and structure of biofilms collected from four full-scale European cooling systems. Water samples were also analyzed to evaluate the impact of suspended microbes on biofilm formation. A common core microbiome, containing members of the families Sphingomonadaceae, Comamonadaceae and Hyphomicrobiaceae, was found in all four biofilms, despite the water of each coming from different sources (river and groundwater). This suggests that selection of the pioneer community was influenced by abiotic factors (temperature, pH) and tolerances to biocides. Members of the Sphingomonadaceae were assumed to play a key role in initial biofilm formation. Subsequent biofilm development was driven primarily by light availability, since biofilms were dominated by phototrophs in the two studied 'open' systems. Their interactions with other microbial populations then shaped the structure of the mature biofilm communities analyzed.

  19. Design evaluation of emergency core cooling systems using Axiomatic Design

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Gyunyoung [Massachusetts Institute of Technology, Department of Mechanical Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)]. E-mail: gheo@mit.edu; Lee, Song Kyu [Korea Advanced Institute of Science and Technology, Department of Nuclear and Quantum Engineering, 373-1 Guseong-dong, Yuseong-gu, Daejeon (Korea, Republic of)

    2007-01-15

    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.

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

    Science.gov (United States)

    2012-06-15

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

  1. Availability analysis of the AP600 passive core cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Syarip, M [National Atomic Energy Research Agency, Yogyakarta (Indonesia); Subki, I R [BATAN Head Office, Jakarta (Indonesia); Canton, M H [Westinghouse Electric Corp. (United States)

    1996-12-01

    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; 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). 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 figs, 3 tabs.

  2. The effects of aging on BWR core isolation cooling systems

    International Nuclear Information System (INIS)

    Lee, B.S.

    1994-10-01

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling (RCIC) system in commercial Boiling Water Reactors (BWRs). This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. 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 failures causes. Current inspection, surveillance, and monitoring practices were also reviewed

  3. Operation method and operation control device for emergency core cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Kinoshita, Shoichiro; Takahashi, Toshiyuki; Fujii, Tadashi [Hitachi Ltd., Tokyo (Japan); Mizutani, Akira

    1996-05-07

    The present invention provides a method of reducing continuous load capacity of an emergency cooling system of a BWR type reactor and a device reducing a rated capacity of an emergency power source facility. Namely, the emergency core cooling system comprises a first cooling system having a plurality of power source systems based on a plurality of emergency power sources and a second cooling system having a remaining heat removing function. In this case, when the first cooling system is operated the manual starting under a predetermined condition that an external power source loss event should occur, a power source division different from the first cooling system shares the operation to operate the secondary cooling system simultaneously. Further, the first cooling system is constituted as a high pressure reactor core water injection system and the second cooling system is constituted as a remaining heat removing system. With such a constitution, a high pressure reactor core water injection system for manual starting and a remaining heat removing system of different power source division can be operated simultaneously before automatic operation of the emergency core cooling system upon loss of external power source of a nuclear power plant. (I.S.)

  4. Emergency core cooling system for LMFBR type reactors

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Fukutomi, Shigeki.

    1980-01-01

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

  5. Emergency core cooling system for a fast reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  6. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  7. Reactor core cooling device

    International Nuclear Information System (INIS)

    Kobayashi, Masahiro.

    1986-01-01

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

  8. Emergency core cooling device

    International Nuclear Information System (INIS)

    Suzaki, Kiyoshi; Inoue, Akihiro.

    1979-01-01

    Purpose: To improve core cooling effect by making the operation region for a plurality of water injection pumps more broader. Constitution: An emergency reactor core cooling device actuated upon failure of recycling pipe ways is adapted to be fed with cooling water through a thermal sleeve by way of a plurality of water injection pump from pool water in a condensate storage tank and a pressure suppression chamber as water feed source. Exhaust pipes and suction pipes of each of the pumps are connected by way of switching valves and the valves are switched so that the pumps are set to a series operation if the pressure in the pressure vessel is high and the pumps are set to a parallel operation if the pressure in the pressure vessel is low. (Furukawa, Y.)

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

    International Nuclear Information System (INIS)

    Watanabe, Yamato; Tazai, Ayuko; Yamagishi, Shohei; Muramatsu, Ken; Muta, Hitoshi

    2014-01-01

    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)

  10. observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

    International Nuclear Information System (INIS)

    Siry, S.A K.

    2007-01-01

    analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

  11. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Sobajima, Makoto

    1985-08-01

    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)

  13. Analysis and prevention of water hammer for the emergency core cooling system

    International Nuclear Information System (INIS)

    Zhao Jun

    2008-01-01

    Emergency core cooling system (ECCS) is an engineered safety feature of nuclear power plant. If the water hammer happens during ECCS injection, the piping system may be broken. It will cause loss of ECC system and affect the safety of reactor core. Based on the functions and characteristics of ECCS and the theory of water hammer, the paper analyzed the potential risk of water hammer in ECCS in Qinshan III, and proposed modifications to prevent the water-hammer damage during ECCS injection. (authors)

  14. Analysis of loss of coolant accident and emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

    1977-01-01

    In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

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

    International Nuclear Information System (INIS)

    Huerta B, A.

    1991-01-01

    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

  16. Power distribution monitoring system in the boiling water cooled reactor core

    International Nuclear Information System (INIS)

    Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

    1987-01-01

    Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

  17. Safety analysis of RSG-GAS Silicide core using one line cooling system

    International Nuclear Information System (INIS)

    Endiah-Puji-Hastuti

    2003-01-01

    In the frame of minimizing the operation-cost, operation mode using one line cooling system is being evaluated. Maximum reactor has been determined and to continuing this program, steady state and transient analysis were done. The analysis was done by means of a core thermal hydraulic code, COOLOD-N, and PARET. The codes solves core thermal hydraulic equation at steady state conditions and transient, respectively. By using silicide core data and coast down flow rate as the input, thermal hydraulics parameters such as fuel cladding and fuel meat temperatures as well as safety margin against flow instability were calculated. Imposing the safety criteria to the results of steady state and transient analysis, maximum permissible power for this operation was obtained as much as 17.1 MW

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

    International Nuclear Information System (INIS)

    Lee, Bom Soon.

    1994-01-01

    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

  19. Reassessment of debris ingestion effects on emergency core cooling-system pump performance

    International Nuclear Information System (INIS)

    Sciacca, F.W.; Rao, D.V.

    2004-01-01

    A study sponsored by the United States (US) Nuclear Regulatory Commission (NRC) was performed to reassess the effects of ingesting loss of coolant accident (LOCA) generated materials into emergency core cooling system (ECCS) pumps and the subsequent impact of this debris on the pumps' ability to provide long-term cooling to the reactor core. ECCS intake systems have been designed to screen out large post-LOCA debris materials. However, small-sized debris can penetrate these intake strainers or screens and reach critical pump components. Prior NRC-sponsored evaluations of possible debris and gas ingestion into ECCS pumps and attendant impacts on pump performance were performed in the early 1980's. The earlier study focused primarily on pressurised water reactor (PWR) ECCS pumps. This issue was revisited both to factor in our improved knowledge of LOCA generated debris and to address specifically both boiling water reactor (BWR) and PWR ECCS pumps. This study discusses the potential effects of ingested debris on pump seals, bearing assemblies, cyclone debris separators, and seal cooling water subsystems. This assessment included both near-term (less than one hour) and long-term (greater than one hour) effects introduced by the postulated LOCA. The work reported herein was performed during 1996-1997. (authors)

  20. Analysis of expediency to set regulators of high-pressure emergency core cooling system of WWER 1000 (B-320)

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Komarov, Yu.A.; Tikhonova, G.G.; Nikiforov, S.N.; Bogodist, V.V.; Fol'tov, I.M.; Khadzh Faradzhallakh Dabbakh, A.

    2011-01-01

    The work shows that setting regulative valves in high-pressure emergency core cooling system of WWER 1000/B-320 can be effective only involving the additional tuning to account traverse speed of operating elements of regulator and configuration of the systems providing cooling of primary loop.

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

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    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)

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

    International Nuclear Information System (INIS)

    Lee, Seungchan; Yoon, Dukjoo; Ha, Sangjun

    2013-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seungchan; Yoon, Dukjoo; Ha, Sangjun [Korea Hydro Nuclear Power Co. Ltd, Daejeon (Korea, Republic of)

    2013-05-15

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

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

    International Nuclear Information System (INIS)

    Edwards, M.K.; Qiu, L.; Guzonas, D.A.

    2010-01-01

    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)

  6. Nuclear reactor core support incorporating also a cooling fluid flow system

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1975-01-01

    A description is given of a core bearing plate with several modular intake units having cooling fluid intake openings on their lower extensions, and on their upper ends located above the bearing plate, at least one fuel assembly which is thus in communication with the area under the bearing plate through the modular intake unit. The means for introducing the cooling fluid into the reactor vessel area are located under the bearing plate. The lower ends of the modular intake have ribs arranged essentially on a plane and join together with openings provided between the seals, in such a manner that the ribs form a barrier. The cooling fluid intake openings are located above this barrier, so that the cooling fluid is compelled to cross it before penetrating into the modular intake units [fr

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

    International Nuclear Information System (INIS)

    Gallardo Padilla, Alberto; Moreyra, Geraldo Lazaro; Nieto Malpartida, Manuel

    2002-01-01

    For design of the Emergency Core Cooling System (ECCS) of reactor RP-10 from Peru is very important the hydraulic analysis of this system. In this paper, based on a basic design of the ECCS are showed the conservation equations, the parabolic movement, being deduced from them the equations to evaluate regarding the time the variables to consider in the design: level of the emergency water in the reserve tank, flow, reaches of sprinkle, etc. In this analysis is considered a quasi-stationary flow for simplify the calculation. The developed model was implemented in a computer program denominated ECCSRP10, in language Fortran 77, whose results are shown in form graph. From analysis of results we can conclude that for the system of pipe of the ECCS the appropriate diameter is of 2 , and that the maximum flow possible to give is of 5 m 3 /h for to assure a minimum time of refrigeration of 150000 seconds. Experimental tests were made in a prototype of the pipe system being demonstrated that the obtained results of the simplified calculation agree with the values registered with a global approach of 10%. (author)

  8. Data Acquisition System Design for Advanced Core-Cooling Mechanism Experiment

    International Nuclear Information System (INIS)

    Zhang, Ziyang; Tian, Fang; Zhang, Tao; Wang, Shen

    2011-01-01

    Data Acquisition System (DAS) design for Advanced Core-Cooling Mechanism Experiment(ACME) is studied in the paper. DAS is an important connection between test facility and result analysis. Firstly, it introduces DAS and its design requirement for ACME. Nearly one thousand data resources need record in ACME. They have different types and acquisition frequencies. In order to record these data, a large scale and high speed layered data acquisition system is developed. Secondly, it discusses the DAS design for ACME, including the analog signal adjusting circuits, clock circuit design, sampling frequencies, data storage and transmission by large database system, anti-interference and etc. Analog signal adjusting circuits are necessary to deal with different kinds of input data to gain standard data resources. Some data change slowly and others change in several seconds according to the test performed on ACME. So it is difficult to use uniform sampling frequencies, and a layered data acquisition system is introduced. A large database is built to store data for ACME test, which keeps data safer and makes subsequent data handling more convenient. A database hot backup is also applied to ensure data safety. The software of DAS is built by Labview, which can provide intuitionist result and friendly interface. Another important function of DAS is the ACME safety protection. Finally, the characteristics and improvement of DAS for ACME is analyzed compared to other test facility. Besides friendly user interface, DAS of ACME can also assure higher data precision and sampling frequency

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

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1986-01-01

    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 [fr

  10. Technical specification improvements to containment heat removal and emergency core cooling systems: Final report

    International Nuclear Information System (INIS)

    Sullivan, W.P.; Ha, C.; Pentzien, D.C.; Visweswaran, S.

    1988-07-01

    This report presents the results of an analysis for technical specification improvements to the emergency core cooling systems (ECCS) and containment heat removal systems (EPRI Research Project 2142-3). The objective of this project is to further develop a reliability- and risk-based methodology to provide improvements by considering groups of surveillance test intervals and allowed out-of-service times jointly. This was done for the technical specifications for the ECCS, containment heat removal equipment, and supporting systems of a boiling water reactor plant. The project (1) developed a methodology for optimizing groups of surveillance test intervals and allowed out-of-service times jointly, (2) applied the methodology in a case study of a specific operating plant, Hatch-2, and (3) evaluated benefits of the application. The results of the case study demonstrate that beneficial technical specification improvements can be realized with application of the methodology. By tightening a small group of sensitive surveillance test intervals (STIs) and allowed out-of-service times (AOTs), a larger group of less sensitive STIs and AOTs can be extended resulting in an overall plant operating cost improvement without reducing the plant safety. The reliability- and risk-based methodology and results from this project can be effectively applied for technical specification improvements at other operating plants

  11. Analysis of the dynamic behaviour of the low-pressure emergency core cooling system tank at Paks NPP

    International Nuclear Information System (INIS)

    1999-01-01

    The low pressure emergency core cooling system tanks (LP ECCS) at WWER-440/V213 units have unique worm-shaped geometry. Analytical and experimental investigations were performed to make an adequate basis for seismic assessment of the worm-shaped tank. The full scale dynamic tests results are presented in comparison with shaking table model experiments and analytical studies. (author)

  12. Analysis of the dynamic behaviour of the low pressure emergency core cooling system tank at Paks NPP

    International Nuclear Information System (INIS)

    Tamas, K.

    2001-01-01

    The low pressure emergency core cooling system tanks (LP ECCS) at WWER-440/V213 units have unique worm-shaped geometry. Analytical and experimental investigations were performed to make an adequate basis for seismic assessment of the worm-shaped tank. The full scale dynamic tests results are presented in comparison with shaking table model experiments and analytical studies. (author)

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

    International Nuclear Information System (INIS)

    Choi, Y.A.; Feltus, M.A.

    1995-01-01

    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

  14. Industry Application Emergency Core Cooling System Cladding Acceptance Criteria Early Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert W. [FPoliSolutions LLC, Murrysville, PA (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Frepoli, Cesare [FPoliSolutions LLC, Murrysville, PA (United States); Yurko, Joseph P. [FPoliSolutions LLC, Murrysville, PA (United States); Swindlehurst, Gregg [GS Nuclear Consulting, Charlotte, NC (United States); Zoino, Angelo [Univ. of Rome Tor Vergata (Italy)

    2015-09-01

    The U. S. NRC is currently proposing rulemaking designated as “10 CFR 50.46c” to revise the loss-of-coolant-accident (LOCA)/emergency core cooling system (ECCS) acceptance criteria to include the effects of higher burnup on cladding performance as well as to address other technical issues. The NRC is also currently resolving the public comments with the final rule expected to be issued in April 2016. The impact of the final 50.46c rule on the industry may involve updating of fuel vendor LOCA evaluation models, NRC review and approval, and licensee submittal of new LOCA evaluations or re-analyses and associated technical specification revisions for NRC review and approval. The rule implementation process, both industry and NRC activities, is expected to take 4-6 years following the rule effective date. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The proposed rule would apply to a light water reactor and to all cladding types.

  15. Emergency reactor core cooling facility

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  16. Emergency reactor core cooling facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

    1996-11-01

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

  17. Concept Design of a Gravity Core Cooling Tank as a Passive Residual Heat Removal System for a Research Reactor

    International Nuclear Information System (INIS)

    Lee, Kwonyeong; Chi, Daeyoung; Kim, Seong Hoon; Seo, Kyoungwoo; Yoon, Juhyeon

    2014-01-01

    A core downward flow is considered to use a plate type fuel because it is benefit to install the fuel in the core. If a flow inversion from a downward to upward flow in the core by a natural circulation is introduced within a high heat flux region of residual heat, the fuel fails instantly due to zero flow. Therefore, the core downward flow should be sufficiently maintained until the residual heat is in a low heat flux region. In a small power research reactor, inertia generated by a flywheel of the PCP can maintain a downward flow shortly and resolve the problem of a flow inversion. However, a high power research reactor more than 10 MW should have an additional method to have a longer downward flow until a low heat flux. Usually, other research reactors have selected an active residual heat removal system as a safety class. But, an active safety system is difficult to design and expensive to construct. A Gravity Core Cooling Tank (GCCT) beside the reactor pool with a Residual Heat Removal Pipe connecting two pools was developed and designed preliminarily as a passive residual heat removal system for an open-pool type research reactor. It is very simple to design and cheap to construct. Additionally, a non-safety, but active residual heat removal system is applied with the GCCT. It is a Pool Water Cooling and Purification System. It can improve the usability of the research reactor by removing the thermal waves, and purify the reactor pool, the Primary Cooling System, and the GCCT. Moreover, it can reduce the pool top radiation level

  18. Improving safety margin of LWRs by rethinking the emergency core cooling system criteria and safety system capacity

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Bokyung, E-mail: bkkim2@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2016-10-15

    Highlights: • Zircaloy embrittlement criteria can increase to 1370 °C for CP-ECR lower than 13%. • The draft ECCS criteria of U.S. NRC allow less than 5% in power margin. • The Japanese fracture-based criteria allow around 5% in power margin. • Increasing SIT inventory is effective in assuring safety margin for power uprates. - Abstract: This study investigates the engineering compatibility between emergency core cooling system criteria and safety water injection systems, in the pursuit of safety margin increase of light water reactors. This study proposes an acceptable temperature increase to 1370 °C as long as equivalent cladding reacted calculated by the Cathcart–Pawel equation is below 13%, after an extensive literature review. The influence of different ECCS criteria on the safety margin during large break loss of coolant accident is investigated for OPR-1000 by the system code MARS-KS, implemented with the KINS-REM method. The fracture-based emergency core cooling system (ECCS) criteria proposed in this study are shown to enable power margins up to 10%. In the meantime, the draft U.S. NRC’s embrittlement criteria (burnup-sensitive) and Japanese fracture-based criteria are shown to allow less than 5%, and around 5% of power margins, respectively. Increasing safety injection tank (SIT) water inventory is the key, yet convenient, way of assuring safety margin for power increase. More than 20% increase in the SIT water inventory is required to allow 15% power margins, for the U.S. NRC’s burnup-dependent embrittlement criteria. Controlling SIT water inventory would be a useful option that could allow the industrial desire to pursue power margins even under the recent atmosphere of imposing stricter ECCS criteria for the considerable burnup effects.

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

    Energy Technology Data Exchange (ETDEWEB)

    Torres, Walmir Maximo; Baptista Filho, Benedito Dias; Ting, Daniel Kao Sun [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Dept. de Tecnologia de Reatores]. E-mail: wmtorres@net.ipen.br; bdbfilho@net.ipen.br; dksting@net.ipen.br

    1998-07-01

    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)

  20. Diversified emergency core cooling in CANDU with a passive moderator heat rejection system

    Energy Technology Data Exchange (ETDEWEB)

    Spinks, N [AECL Research, Chalk River Labs., Chalk River, ON (Canada)

    1996-12-01

    A passive moderator heat rejection system is being developed for CANDU reactors which, combined with a conventional emergency-coolant injection system, provides the diversity to reduce core-melt frequency to order 10{sup -7} per unit-year. This is similar to the approach used in the design of contemporary CANDU shutdown systems which leads to a frequency of order 10{sup -8} per unit-year for events leading to loss of shutdown. Testing of a full height 1/60 power-and-volume-scaled loop has demonstrated the feasibility of the passive system for removal of moderator heat during normal operation and during accidents. With the frequency of core-melt reduced, by these measures, to order 10{sup -7} per unit year, no need should exist for further mitigation. (author). 3 refs, 2 figs.

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    1988-12-01

    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

  3. A Bayesian reliability study on motorized valves for the emergency core cooling, heat transport isolation and shutdown cooling systems at Gentilly-2 Nuclear Generating Station

    International Nuclear Information System (INIS)

    Smith, J.E.; Rennick, D.F.; Nainer, A.

    1996-01-01

    The objective of this is to examine operational data on 32 motorized valves in the emergency core cooling, shutdown cooling and heat transport isolation systems and determine if the evidence would support a reduction in testing frequency of these valves. The methodology used is to examine the data which has accumulated on motorized valve failures since Gentilly-2 first entered service, compare these data with similar data from other sources, and determine whether the evidence indicate that demand-based, wear out type failure mechanisms play a significant role in the recorded failures. The statistical data are then updated, using a Bayesian updating procedure, to obtain revised time based failure rates and demand based probabilities of failure on demand for the motorized valves. The revised failure rates and probabilities are then applied to the fault tree models for the systems of interest to determine what effects there would be, with the current test intervals and with extended test intervals, on the probability of failure of the systems. (author)

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

    International Nuclear Information System (INIS)

    Chupin, A.; Hu, L. W.; Buongiorno, J.

    2008-01-01

    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)

  5. Failure Mode and Effects Analysis (FMEA) of the Emergency Core Cooling System (ECCS) for a Westinghouse type 312, three loop pressurized water reactor

    International Nuclear Information System (INIS)

    Shopsky, W.E.

    1977-01-01

    The Emergency Core Cooling System (ECCS) is a Safeguards System designed to cool the core in the unlikely event of a Loss-of-Coolant Accident (LOCA) in the primary reactor coolant system as well as to provide additional shutdown capability following a steam break accident. The system is designed for a high reliability of providing emergency coolant and shutdown reactivity to the core for all anticipated occurrences of such accidents. The ECCS by performing its intended function assures that fuel and clad damage is minimized during accident conditions thus reducing release of fission products from the fuel. The ECCS is designed to perform its function despite sustaining a single failure by the judicious use of equipment and flow path redundancy within and outside the containment structure. By the use of an analytic tool, a Failure Mode and Effects Analysis (FMEA), it is shown that the ECCS is in compliance with the Single Failure Criterion established for active failures of fluid systems during short and long term cooling of the reactor core following a LOCA or steam break accident. An analysis was also performed with regards to passive failure of ECCS components during long-term cooling of the core following an accident. The design of the ECCS was verified as being able to tolerate a single passive failure during long-term cooling of the reactor core following an accident. The FMEA conducted qualitatively demonstrates the reliability of the ECCS (concerning active components) to perform its intended safety function

  6. Experimental analysis of ex-vessel core catcher cooling system performance for EU-APR1400 during severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Song, K. W.; Park, H. S.; Revankar, S. T. [POSTECH, Pohang (Korea, Republic of); Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In the coolant channel which has a unique design and large scale flow paths, natural circulation is passively activated by buoyancy driven force. Since two-phase flow behavior in a large scale channel is different from that in a small scale channel, the two-phase flow affecting the cooling capability is difficult to be predicted in the large channel. Therefore, cooling experiment in the core catcher coolant path is necessary. Cooling Experiment - Passive Ex-vessel corium retaining and Cooling System(CE-PECS) is constructed in full scale(in height and width) slice of half prototype. It actually simulates steam-water flow in the coolant channel for different decay heat condition of the corium. In this study, thermal power considering of total amount of decay heat 190 kW which corresponds to 40MW of thermal power in the prototype is loaded on the top wall of the CE-PECS coolant channel. Natural circulation flow rate and pressure drops at the two-phase region are measured in various power level. Temperatures of heater block and working fluid in various position along the flow path enable to calculate heat fluxes and heat transfer coefficients distribution. These results are used for evaluating heat removal capability of core catcher facility. Two-phase natural circulation experiment is carried out in CE-PECS facility. Based on the prototypic condition, 190 kW of total power is supplied to the top of the coolant path. Uniform distribution of heat load on the downward facing heater bock produces -300 kW/m2 at 100 % power ratio. Although the experiment should consider the heat loss and heat flux uniformity, several noticeable conclusions have been made as followings; 1. Mass flow rate and two-phase pressure drop are measured in various power conditions. 2. Slightly inclined top wall at the downstream of the channel shows better heat exchange performance than horizontal top wall because enhanced convection due to the increase of void fraction improves local cooling. This

  7. Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

    CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

  8. Cooling systems

    International Nuclear Information System (INIS)

    Coutant, C.C.

    1978-01-01

    Progress on the thermal effects project is reported with regard to physiology and distribution of Corbicula; power plant effects studies on burrowing mayfly populations; comparative thermal responses of largemouth bass from northern and southern populations; temperature selection by striped bass in Cherokee Reservoir; fish population studies; and predictive thermoregulation by fishes. Progress is also reported on the following; cause and ecological ramifications of threadfin shad impingement; entrainment project; aquaculture project; pathogenic amoeba project; and cooling tower drift project

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

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Baptista Filho, Benedito Dias; Ting, Daniel Kao Sun

    1999-01-01

    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)

  10. Effects of debris generated by chemical reactions on head loss through emergency-core cooling-system strainers

    International Nuclear Information System (INIS)

    Howe, K.; Ghosh, A.; Maji, A.K.; Letellier, B.C.; Johns, R.; Chang, T.Y.

    2004-01-01

    The effect of debris generated during a loss of coolant accident (LOCA) on the emergency core cooling system (ECCS) strainers has been studied via numerous avenues over the last several years. The research described in this manuscript examines the generation and effect of secondary materials -- not debris generated in the LOCA itself, but materials created by chemical reactions between exposed surfaces/debris and cooling system water. The secondary materials studied in the research were corrosion products from exposed metallic surfaces and paint chips that may precipitate out of solution, with a focus on the corrosion products of aluminium, iron, and zinc. The processes of corrosion and leaching of metals with subsequent precipitation is important because: (1) the surface area of exposed metal inside containment represents a large potential source term, even for slow chemical reactions; the chemical composition of the cooling system water (boric acid, lithium, etc.) may affect corrosion or precipitation in ways that have not been studied thoroughly in the past; and (3) an eyewitness report of the presence of gelatinous material in the Three Mile Island containment pool after the 1979 accident suggests the formation of a secondary material that has not been examined under the generic safety issue (GSI)-191 research program. This research was limited in scope and consisted only of small-scale tests. Several key questions were investigated: (1) do credible corrosion mechanisms exist for leaching metal ions from bulk solid surfaces or from zinc-based paint chips, and if so, what are the typical rate constants? (2) can corrosion products accumulate in the containment pool water to the extent that they might precipitate as new chemical species at pH and temperatures levels that are relevant to the LOCA accident sequence? and (3) how do chemical precipitants affect the head loss across an existing fibrous debris bed? A full report of the research is available. (authors)

  11. Special power supply and control system for the gas-cooled fast reactor-core flow test loop

    International Nuclear Information System (INIS)

    Hudson, T.L.

    1981-09-01

    The test bundle in the Gas-Cooled Fast Reactor-Core Flow Test Loop (GCFR-CFTL) requires a source of electrical power that can be controlled accurately and reliably over a wide range of steady-state and transient power levels and skewed power distributions to simulate GCFR operating conditions. Both ac and dc power systems were studied, and only those employing silicon-controlled rectifiers (SCRs) could meet the requirements. This report summarizes the studies, tests, evaluations, and development work leading to the selection. it also presents the design, procurement, testing, and evaluation of the first 500-kVa LMPL supply. The results show that the LMPL can control 60-Hz sine wave power from 200 W to 500 kVA

  12. A very cool cooling system

    CERN Multimedia

    Antonella Del Rosso

    2015-01-01

    The NA62 Gigatracker is a jewel of technology: its sensor, which delivers the time of the crossing particles with a precision of less than 200 picoseconds (better than similar LHC detectors), has a cooling system that might become the precursor to a completely new detector technique.   The 115 metre long vacuum tank of the NA62 experiment. The NA62 Gigatracker (GTK) is composed of a set of three innovative silicon pixel detectors, whose job is to measure the arrival time and the position of the incoming beam particles. Installed in the heart of the NA62 detector, the silicon sensors are cooled down (to about -20 degrees Celsius) by a microfluidic silicon device. “The cooling system is needed to remove the heat produced by the readout chips the silicon sensor is bonded to,” explains Alessandro Mapelli, microsystems engineer working in the Physics department. “For the NA62 Gigatracker we have designed a cooling plate on top of which both the silicon sensor and the...

  13. The application of redundancy-related basic safety principles to the 1400 MWE reactor core standby cooling system

    International Nuclear Information System (INIS)

    Bertrand, R.

    1990-01-01

    This memorandum shall provide the background for the work of the European Community Commission which is to analyze safety principles relating to redundancy. The redundancy-related basic safety principles applied in French nuclear power plants are the following: . the single-failure criterion, . provisions additional to application of the single-failure criterion. These are mainly provisions made at the design stage to minimize risks associated with common cause failures or the risks of human error which can lead to such failures: - protection against hazards of internal and external origin, - the geographical or physical separation of equipment, - the independence of electrical power supplies and distribution systems, - the additional resources and associated operating procedures making it possible to accommodate total loss of the safety systems. The scope also includes the operating rules which ensure availability of redundant safety-related equipment. The provisions relating to the single-failure criterion are detailed in Basic Safety Rule 1.3.A appended. The application of these principles proposed by the operating organization and accepted by the safety authorities for the design and operation of the standby core cooling system (System RIS) is explained

  14. The use of reliability analysis techniques applied to nuclear power station emergency core cooling systems

    International Nuclear Information System (INIS)

    Danielsen, A.; Snaith, E.R.

    1975-01-01

    A reliability investigation carried out by the Safety and Reliability Services of the UKAEA, and the SSEB, of the essential system/reactor coolant system for a large nuclear power station is described. In AGR type reactors, after all reactor shutdown conditions, it is necessary to restore forced gas circulation and sufficient boiler feed to maintain the heat removal capacity of the boilers. The coolant requirements are provided by several independent mechanical systems of primary coolant fans, feedwater pumps, and valves integrated with electrical power sources, switchgear, and automatic control equipment. Reliability is treated as one aspect of system performance and quantified in terms of failure to meet a specific objective. Based on the reliability performance of the constituent components the optimum system configuration is determined together with the preferred plant operating procedures and maintenance requirements. (author)

  15. Heat Removal Performance of Hybrid Control Rod for Passive In-Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    The two-phase closed heat transfer device can be divided by thermosyphon heat pipe and capillary wicked heat pipe which uses gravitational force or capillary pumping pressure as a driving force of the convection of working fluid. If there is a temperature difference between reactor core and ultimate heat sink, the decay heat removal and reactor shutdown is possible at any accident conditions without external power sources. To apply the hybrid control rod to the commercial nuclear power plants, its modelling about various parameters is the most important work. Also, its unique geometry is coexistence of neutron absorber material and working fluid in a cladding material having annular vapor path. Although thermosyphon heat pipe (THP) or wicked heat pipe (WHP) shows high heat transfer coefficients for limited space, the maximum heat removal capacity is restricted by several phenomena due to their unique heat transfer mechanism. Validation of the existing correlations on the annular vapor path thermosyphon (ATHP) which has different wetted perimeter and heated diameter must be conducted. The effect of inner structure, and fill ratio of the working fluid on the thermal performance of heat pipe has not been investigated. As a first step of the development of hybrid heat pipe, the ATHP which contains neutron absorber in the concentric thermosyphon (CTHP) was prepared and the thermal performance of the annular thermosyphon was experimentally studied. The heat transfer characteristics and flooding limit of the annular vapor path thermosyphon was studied experimentally to model the performance of hybrid control rod. The following results were obtained: (1) The annular vapor path thermosyphon showed better evaporation heat transfer due to the enhanced convection between adiabatic and condenser section. (2) Effect of fill ratio on the heat transfer characteristics was negligible. (3) Existing correlations about flooding limit of thermosyphon could not reflect the annular vapor

  16. Process fluid cooling system

    International Nuclear Information System (INIS)

    Farquhar, N.G.; Schwab, J.A.

    1977-01-01

    A system of heat exchangers is disclosed for cooling process fluids. The system is particularly applicable to cooling steam generator blowdown fluid in a nuclear plant prior to chemical purification of the fluid in which it minimizes the potential of boiling of the plant cooling water which cools the blowdown fluid

  17. The development of test software for the inadequate core cooling monitoring system

    International Nuclear Information System (INIS)

    Lee, Soon Sung.

    1996-06-01

    The test software including the ICCMS simulator which is necessary for dynamic test for the ICCMS software in PWR is developed. The developed dynamic test software consists of the module test simulator, the integration test simulator, and the test result analyser. The simulator was programmed by C language according to the same algorithm requirements for the FORTRAN version ICCMS software, and also for the Factory Acceptance Test (FAT). And the simulator can be used as training tool for the reactor operator and system development tool for the performance improvement. (author). 4 tabs., 8 figs., 11 refs

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

    International Nuclear Information System (INIS)

    Muncharoen, C.; Chanyotha, S.; Bereznai, G.

    1998-01-01

    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

  19. A study on design improvement of the emergency core cooling system for a nuclear ship reactor

    International Nuclear Information System (INIS)

    Naruko, Yoshinori

    1982-01-01

    ECCS performances are predicted for the N.S. ''MUTSU'' Reactor using new computing techniques. The actual performances are found to be inadequate with regard to the total injection flow rate and the infection head of the High Pressure Injection System (HPIS). The actual Safety Injection (SI) Signal is also shown to be inoperative in a gas phase LOCA. Because of this fact, the ECCS has been improved by replacing the existing HPIS with two large high pressure pumps and by adding ''Reactor Pressure Low-Low'' and ''Containment Pressure High'' signals. This report deals with a numerical study of the dynamical behavior of the N.S. MUTSU Reactor in the postulated small break LOCA, which has been calculated to verify the design improvement by evaluating the new ECCS performances on the basis of RELAP4/MOD6/SUS and TOODEE2/SUS codes. In conclusion, the improved system performs satisfactorily in the whole range of gas phase breaks and in the small break range of liquid phase LOCA. (author)

  20. Cooled Water Production System,

    Science.gov (United States)

    The invention refers to the field of air conditioning and regards an apparatus for obtaining cooled water . The purpose of the invention is to develop...such a system for obtaining cooled water which would permit the maximum use of the cooling effect of the water -cooling tower.

  1. Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Murao, Yoshio

    1985-01-01

    In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15 K using the one-dimensional model in the CCTF tests. The CCTF results support the one-dimensional treatment assumed in the reactor safety assessment. (author)

  2. Gravity driven emergency core cooling experiments with the PACTEL facility

    International Nuclear Information System (INIS)

    Munther, R.; Kalli, H.; Kouhia, J.

    1996-01-01

    PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs

  3. Gravity driven emergency core cooling experiments with the PACTEL facility

    Energy Technology Data Exchange (ETDEWEB)

    Munther, R; Kalli, H [University of Technology, Lappeenranta (Finland); Kouhia, J [Technical Research Centre of Finland, Lappeenranta (Finland)

    1996-12-01

    PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs.

  4. Warming rays in cluster cool cores

    Science.gov (United States)

    Colafrancesco, S.; Marchegiani, P.

    2008-06-01

    Context: Cosmic rays are confined in the atmospheres of galaxy clusters and, therefore, they can play a crucial role in the heating of their cool cores. Aims: We discuss here the thermal and non-thermal features of a model of cosmic ray heating of cluster cores that can provide a solution to the cooling-flow problems. To this aim, we generalize a model originally proposed by Colafrancesco, Dar & DeRujula (2004) and we show that our model predicts specific correlations between the thermal and non-thermal properties of galaxy clusters and enables various observational tests. Methods: The model reproduces the observed temperature distribution in clusters by using an energy balance condition in which the X-ray energy emitted by clusters is supplied, in a quasi-steady state, by the hadronic cosmic rays, which act as “warming rays” (WRs). The temperature profile of the intracluster (IC) gas is strictly correlated with the pressure distribution of the WRs and, consequently, with the non-thermal emission (radio, hard X-ray and gamma-ray) induced by the interaction of the WRs with the IC gas and the IC magnetic field. Results: The temperature distribution of the IC gas in both cool-core and non cool-core clusters is successfully predicted from the measured IC plasma density distribution. Under this contraint, the WR model is also able to reproduce the thermal and non-thermal pressure distribution in clusters, as well as their radial entropy distribution, as shown by the analysis of three clusters studied in detail: Perseus, A2199 and Hydra. The WR model provides other observable features of galaxy clusters: a correlation of the pressure ratio (WRs to thermal IC gas) with the inner cluster temperature (P_WR/P_th) ˜ (kT_inner)-2/3, a correlation of the gamma-ray luminosity with the inner cluster temperature Lγ ˜ (kT_inner)4/3, a substantial number of cool-core clusters observable with the GLAST-LAT experiment, a surface brightness of radio halos in cool-core clusters

  5. Radiant Floor Cooling Systems

    DEFF Research Database (Denmark)

    Olesen, Bjarne W.

    2008-01-01

    In many countries, hydronic radiant floor systems are widely used for heating all types of buildings such as residential, churches, gymnasiums, hospitals, hangars, storage buildings, industrial buildings, and smaller offices. However, few systems are used for cooling.This article describes a floor...... cooling system that includes such considerations as thermal comfort of the occupants, which design parameters will influence the cooling capacity and how the system should be controlled. Examples of applications are presented....

  6. Testing Numerical Models of Cool Core Galaxy Cluster Formation with X-Ray Observations

    Science.gov (United States)

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

    2009-12-01

    Using archival Chandra and ROSAT data along with numerical simulations, we compare the properties of cool core and non-cool core galaxy clusters, paying particular attention to the region beyond the cluster cores. With the use of single and double β-models, we demonstrate a statistically significant difference in the slopes of observed cluster surface brightness profiles while the cluster cores remain indistinguishable between the two cluster types. Additionally, through the use of hardness ratio profiles, we find evidence suggesting cool core clusters are cooler beyond their cores than non-cool core clusters of comparable mass and temperature, both in observed and simulated clusters. The similarities between real and simulated clusters supports a model presented in earlier work by the authors describing differing merger histories between cool core and non-cool core clusters. Discrepancies between real and simulated clusters will inform upcoming numerical models and simulations as to new ways to incorporate feedback in these systems.

  7. Cooling water distribution system

    Science.gov (United States)

    Orr, Richard

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

  8. Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, Aaron [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Vaghetto, Rodolfo [Texas A & M Univ., College Station, TX (United States); Vanni, Alessandro [Texas A & M Univ., College Station, TX (United States); Neptune, Kaleb [Texas A & M Univ., College Station, TX (United States)

    2017-06-01

    Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

  9. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    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) [pt

  10. Experimental study on two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400 using air-water system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Won [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Nguyen, Thanh Hung [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Ha, Kwang Soon; Kim, Hwan Yeol; Song, Jinho [Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Park, Hyun Sun [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Revankar, Shripad T., E-mail: shripad@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); School of Nuclear Engineering, Purdue University, West Lafayette, IN 47906 (United States); Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Institute of Nuclear Safety, Daejeon 305-338 (Korea, Republic of)

    2017-05-15

    Highlights: • Two-phase flow regimes and transition behavior were observed in the coolant channel. • Test were conducted for natural circulation with air-water. • Data were obtained on flow regime, void fraction, flow rates and re-wetting time. • The data were related to a cooling capability of core catcher system. - Abstract: Ex-vessel core catcher cooling system driven by natural circulation is designed using a full scaled air-water system. A transparent half symmetric section of a core catcher coolant channel of a pressurized water reactor was designed with instrumentations for local void fraction measurement and flow visualization. Two designs of air-water top separator water tanks are studied including one with modified ‘super-step’ design which prevents gas entrainment into down-comer. In the experiment air flow rates are set corresponding to steam generation rate for given corium decay power. Measurements of natural circulation flow rate, spatial local void fraction distribution and re-wetting time near the top wall are carried out for various air flow rates which simulate boiling-induced vapor generation. Since heat transfer and critical heat flux are strongly dependent on the water mass flow rate and development of two-phase flow on the heated wall, knowledge of two-phase flow characteristics in the coolant channel is essential. Results on flow visualization showing two phase flow structure specifically near the high void accumulation regions, local void profiles, rewetting time, and natural circulation flow rate are presented for various air flow rates that simulate corium power levels. The data are useful in assessing the cooling capability of and safety of the core catcher system.

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  12. OSCIL: one-dimensional spring-mass system simulator for seismic analysis of high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Lasker, L.

    1976-01-01

    OSCIL is a program to predict the effects of seismic input on a HTGR core. The present model is a one-dimensional array of blocks with appropriate spring constants, inter-elemental and ground damping, and clearances. It can be used more generally for systems of moving masses separated by nonlinear springs and dampers

  13. OSCIL: one-dimensional spring-mass system simulator for seismic analysis of high temperature gas cooled reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Lasker, L. (ed.)

    1976-01-01

    OSCIL is a program to predict the effects of seismic input on a HTGR core. The present model is a one-dimensional array of blocks with appropriate spring constants, inter-elemental and ground damping, and clearances. It can be used more generally for systems of moving masses separated by nonlinear springs and dampers.

  14. 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)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

    2015-01-01

    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

  15. Monitoring Cray Cooling Systems

    Energy Technology Data Exchange (ETDEWEB)

    Maxwell, Don E [ORNL; Ezell, Matthew A [ORNL; Becklehimer, Jeff [Cray, Inc.; Donovan, Matthew J [ORNL; Layton, Christopher C [ORNL

    2014-01-01

    While sites generally have systems in place to monitor the health of Cray computers themselves, often the cooling systems are ignored until a computer failure requires investigation into the source of the failure. The Liebert XDP units used to cool the Cray XE/XK models as well as the Cray proprietary cooling system used for the Cray XC30 models provide data useful for health monitoring. Unfortunately, this valuable information is often available only to custom solutions not accessible by a center-wide monitoring system or is simply ignored entirely. In this paper, methods and tools used to harvest the monitoring data available are discussed, and the implementation needed to integrate the data into a center-wide monitoring system at the Oak Ridge National Laboratory is provided.

  16. Turbine airfoil cooling system with cooling systems using high and low pressure cooling fluids

    Science.gov (United States)

    Marsh, Jan H.; Messmann, Stephen John; Scribner, Carmen Andrew

    2017-10-25

    A turbine airfoil cooling system including a low pressure cooling system and a high pressure cooling system for a turbine airfoil of a gas turbine engine is disclosed. In at least one embodiment, the low pressure cooling system may be an ambient air cooling system, and the high pressure cooling system may be a compressor bleed air cooling system. In at least one embodiment, the compressor bleed air cooling system in communication with a high pressure subsystem that may be a snubber cooling system positioned within a snubber. A delivery system including a movable air supply tube may be used to separate the low and high pressure cooling subsystems. The delivery system may enable high pressure cooling air to be passed to the snubber cooling system separate from low pressure cooling fluid supplied by the low pressure cooling system to other portions of the turbine airfoil cooling system.

  17. Reactor cooling system

    International Nuclear Information System (INIS)

    Kato, Etsuji.

    1979-01-01

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

  18. Modelling of thermohydraulic emergency core cooling phenomena

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Andreani, M.; Lewis, M.J.

    1990-10-01

    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

  19. ITER cooling systems

    International Nuclear Information System (INIS)

    Natalizio, A.; Hollies, R.E.; Sochaski, R.O.; Stubley, P.H.

    1992-06-01

    The ITER reference system uses low-temperature water for heat removal and high-temperature helium for bake-out. As these systems share common equipment, bake-out cannot be performed until the cooling system is drained and dried, and the reactor cannot be started until the helium has been purged from the cooling system. This study examines the feasibility of using a single high-temperature fluid to perform both heat removal and bake-out. The high temperature required for bake-out would also be in the range for power production. The study examines cost, operational benefits, and impact on reactor safety of two options: a high-pressure water system, and a low-pressure organic system. It was concluded that the cost savings and operational benefits are significant; there are no significant adverse safety impacts from operating either the water system or the organic system; and the capital costs of both systems are comparable

  20. Evaluation of load case ''switch-off of the high pressure pump of the emergency core cooling system'', measures of verification and in situ-test

    International Nuclear Information System (INIS)

    Trobitz, M.; Mattheis, A.; Kerkhof, K.; Hippelein, K.; Hofstoetter, P.

    1998-01-01

    Within the framework of periodic safety inspection of the Gundremmingen power station (RWE-Bayernwerk - KRB II), the load collectives used for the design of safety-relevant systems and components were checked for their consistency with latest updates of the design basis. It was found that there was no analytical information or study available describing a particular process and its effects, namely switch-off of the high-pressure feedwater pump of the emergency core cooling system. The paper reports the work performed for closing the gap, including preparatory analyses, accompanying measures such as vibration measurements during plant shut-down, as well as the preparation and performance of the in-situ test. The experimental results and the comparative evaluation of calculated and experimental data are presented. (orig./CB) [de

  1. Analysis of the RBMK-1500 type reactor emergency core cooling system behavior, taking into account the specified hydraulic characteristics of fast acting motor valves

    International Nuclear Information System (INIS)

    Kaliatka, A.; Ognerubov, V.; Adomavicius, A.; Ziedelis, S.

    2005-01-01

    During the accident analysis of nuclear power plants, reliability and uncertainty of results depends on adequateness of mathematical models of main elements and phenomena in systems important to safety. The best way for qualification of these models is collation with relevant experimental data. However, at the case of lack of such data modern computational fluid dynamics codes can be used for this purpose. This paper presents the results of an attempt to specify the hydraulic characteristics of the fast acting motor valves as well as to demonstrate the impact of these characteristics to transient processes in emergency core cooling system of the RBMK-1500 type reactor. For these purposes the finite element model of fast acting motor valve was developed and analyzed, using two separate computational fluid dynamics codes in parallel: CFX5 and COSMOS/FLOWORKS. Both all main design particularities and changes of flow structure during valve opening (closure) process were taken into account. It was demonstrated, that the obtained dependencies of changes of hydraulic loss coefficient in respect of relative valve opening (closure) rate substantially differ from those commonly used in thermal-hydraulic calculations of nuclear reactors. This difference is extremely big at the square one of the valve opening process, when the value of the valve hydraulic resistance is most important to flow of coolant channelized to the group distribution header. The series of thermal-hydraulic calculations of the maximum design-basis accident initiated by full break of main circulation pump pressure header were performed. The obtained dependencies of changes of hydraulic loss coefficient in respect of relative valve opening (closure) rate as well as those commonly used in thermal-hydraulic code RELAP5 were used. The results of calculations show, that in the initial stage of accident flow of coolant going from emergency core cooling system via fast acting motor valves to group distribution

  2. Simulation of Two-Phase Natural Circulation Loop for Core Cather Cooling Using Air Water

    International Nuclear Information System (INIS)

    Revankar, S. T.; Huang, S. F.; Song, K. W.; Rhee, B. W.; Park, R. J.; Song, J. H.

    2012-01-01

    A closed loop natural circulation system employs thermally induced density gradients in single phase or two-phase liquid form to induce circulation of the working fluid thereby obviating the need for any mechanical moving parts such as pumps and pump controls. This increases the reliability and safety of the cooling system and reduces installation, operation and maintenance costs. That is the reason natural circulation cooling has been considered in advanced reactor core cooling and in engineered safety systems. Natural circulation cooling has been proposed to remove reactor decay heat by external vessel cooling for in-vessel core retention during sever accident scenario. Recently in APR1400 reactor core catcher design natural circulation cooling is proposed to stabilize and cool the corium ejected from the reactor vessel following core melt and breach of reactor vessel. The natural circulation flow is similar to external vessel cooling where water flows through an inclined narrow gap below hot surface and is heated to produce boiling. The two-phase natural circulation enables cooling of the corium pool collected on core catcher. Due to importance of this problem this paper focuses simulation of the two-phase natural circulation through inclined gap using air-water system. Scaling criteria for air-water loop are derived that enable simulation of the flow regimes and natural circulation flow rates in such systems using air-water system

  3. Supplmental testimony of the AEC Regulatory Staff. Public rulemaking hearing on: interim acceptance criteria for emergency core cooling systems for light-water cooled power reactors

    International Nuclear Information System (INIS)

    1972-01-01

    Information is presented concerning sensitivity analysis, loop codes, two-phase pressure drop, critical flow model, pump modeling, PWR core flow distribution, accumulator bypass, fuel densification, gap thermal conductance and UO 2 thermal conductivity, transition boiling heat transfer, clad-to-fluid heat transfer, heat transfer at low pressure, reflood rate analyses, containment back pressure during reflood, BWR FLECHT, PWR reflooding heat transfer FLECHT data, embrittlement and post-blowdown loads, fuel rod physico-chemical reactions, flow blockage, small break analysis, and decay heat. (U.S.)

  4. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

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

  5. 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

  6. 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)

    Lioce, D.; Asztalos, M.; Alemberti, A.; Barucca, L.; Frogheri, M.; Saiu, G.

    2012-01-01

    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

  7. ITER cooling system

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-11-01

    The present specification of the ITER cooling system does not permit its operation with water above 150 C. However, the first wall needs to be heated to higher temperatures during conditioning at 250 C and bake-out at 350 C. In order to use the cooling water for these operations the cooling system would have to operate during conditioning at 37 Bar and during bake-out at 164 Bar. This is undesirable from the safety analysis point of view, and alternative heating methods are to be found. This review suggests that superheated steam or gas heating can be used for both baking and conditioning. The blanket design must consider the use of dual heat transfer media, allowing for change from one to another in both directions. Transfer from water to gas or steam is the most intricate and risky part of the entire heating process. Superheated steam conditioning appears unfavorable. The use of inert gas is recommended, although alternative heating fluids such as organic coolant should be investigated

  8. Hydraulic analysis of emergency core cooling system of reactor RP-10; Analisis hidraulico del sistema de refrigeracion de emergencia del nucleo del reactor RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo Padilla, Alberto; Moreyra, Geraldo Lazaro; Nieto Malpartida, Manuel [Instituto Peruano de Energia Nuclear, Lima (Peru)]. E-mail: agallardo@ipen.gob.pe; glazaro@ipen.gob.pe; mnieto@ipen.gob.pe

    2002-07-01

    For design of the Emergency Core Cooling System (ECCS) of reactor RP-10 from Peru is very important the hydraulic analysis of this system. In this paper, based on a basic design of the ECCS are showed the conservation equations, the parabolic movement, being deduced from them the equations to evaluate regarding the time the variables to consider in the design: level of the emergency water in the reserve tank, flow, reaches of sprinkle, etc. In this analysis is considered a quasi-stationary flow for simplify the calculation. The developed model was implemented in a computer program denominated ECCSRP10, in language Fortran 77, whose results are shown in form graph. From analysis of results we can conclude that for the system of pipe of the ECCS the appropriate diameter is of 2 in., and that the maximum flow possible to give is of 5 m{sup 3}/h for to assure a minimum time of refrigeration of 150000 seconds. Experimental tests were made in a prototype of the pipe system being demonstrated that the obtained results of the simplified calculation agree with the values registered with a global approach of 10%. (author)

  9. 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.

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

    International Nuclear Information System (INIS)

    Agrell, Maria; Sandervag, Oddbjoern; Amri, Abdallah; ); Bang, Young S.; Blomart, Philippe; Broecker, Annette; Pointner, Winfried; Ganzmann, Ingo; Lenogue, Bruno; Guzonas, David; Herer, Christophe; Mattei, Jean-Marie; Tricottet, Matthieu; Masaoka, Hideaki; Soltesz, Vojtech; Tarkiainen, Seppo; Ui, Atsushi; Villalba, Cristina; Zigler, Gilbert

    2013-11-01

    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

  11. STAR FORMATION EFFICIENCY IN THE COOL CORES OF GALAXY CLUSTERS

    International Nuclear Information System (INIS)

    McDonald, Michael; Veilleux, Sylvain; Mushotzky, Richard; Reynolds, Christopher; Rupke, David S. N.

    2011-01-01

    We have assembled a sample of high spatial resolution far-UV (Hubble Space Telescope Advanced Camera for Surveys/Solar Blind Channel) and Hα (Maryland-Magellan Tunable Filter) imaging for 15 cool core galaxy clusters. These data provide a detailed view of the thin, extended filaments in the cores of these clusters. Based on the ratio of the far-UV to Hα luminosity, the UV spectral energy distribution, and the far-UV and Hα morphology, we conclude that the warm, ionized gas in the cluster cores is photoionized by massive, young stars in all but a few (A1991, A2052, A2580) systems. We show that the extended filaments, when considered separately, appear to be star forming in the majority of cases, while the nuclei tend to have slightly lower far-UV luminosity for a given Hα luminosity, suggesting a harder ionization source or higher extinction. We observe a slight offset in the UV/Hα ratio from the expected value for continuous star formation which can be modeled by assuming intrinsic extinction by modest amounts of dust (E(B - V) ∼ 0.2) or a top-heavy initial mass function in the extended filaments. The measured star formation rates vary from ∼0.05 M sun yr -1 in the nuclei of non-cooling systems, consistent with passive, red ellipticals, to ∼5 M sun yr -1 in systems with complex, extended, optical filaments. Comparing the estimates of the star formation rate based on UV, Hα, and infrared luminosities to the spectroscopically determined X-ray cooling rate suggests a star formation efficiency of 14 +18 -8 %. This value represents the time-averaged fraction, by mass, of gas cooling out of the intracluster medium, which turns into stars and agrees well with the global fraction of baryons in stars required by simulations to reproduce the stellar mass function for galaxies. This result provides a new constraint on the efficiency of star formation in accreting systems.

  12. A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

    International Nuclear Information System (INIS)

    Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

    2012-01-01

    A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

  13. Lamination cooling system

    Science.gov (United States)

    Rippel, Wally E.; Kobayashi, Daryl M.

    2005-10-11

    An electric motor, transformer or inductor having a lamination cooling system including a stack of laminations, each defining a plurality of apertures at least partially coincident with apertures of adjacent laminations. The apertures define a plurality of cooling-fluid passageways through the lamination stack, and gaps between the adjacent laminations are sealed to prevent a liquid cooling fluid in the passageways from escaping between the laminations. The gaps are sealed by injecting a heat-cured sealant into the passageways, expelling excess sealant, and heat-curing the lamination stack. The apertures of each lamination can be coincident with the same-sized apertures of adjacent laminations to form straight passageways, or they can vary in size, shape and/or position to form non-axial passageways, angled passageways, bidirectional passageways, and manifold sections of passageways that connect a plurality of different passageway sections. Manifold members adjoin opposite ends of the lamination stack, and each is configured with one or more cavities to act as a manifold to adjacent passageway ends. Complex manifold arrangements can create bidirectional flow in a variety of patterns.

  14. Hybrid radiator cooling system

    Science.gov (United States)

    France, David M.; Smith, David S.; Yu, Wenhua; Routbort, Jules L.

    2016-03-15

    A method and hybrid radiator-cooling apparatus for implementing enhanced radiator-cooling are provided. The hybrid radiator-cooling apparatus includes an air-side finned surface for air cooling; an elongated vertically extending surface extending outwardly from the air-side finned surface on a downstream air-side of the hybrid radiator; and a water supply for selectively providing evaporative cooling with water flow by gravity on the elongated vertically extending surface.

  15. Rotary engine cooling system

    Science.gov (United States)

    Jones, Charles (Inventor); Gigon, Richard M. (Inventor); Blum, Edward J. (Inventor)

    1985-01-01

    A rotary engine has a substantially trochoidal-shaped housing cavity in which a rotor planetates. A cooling system for the engine directs coolant along a single series path consisting of series connected groups of passages. Coolant enters near the intake port, passes downwardly and axially through the cooler regions of the engine, then passes upwardly and axially through the hotter regions. By first flowing through the coolest regions, coolant pressure is reduced, thus reducing the saturation temperature of the coolant and thereby enhancing the nucleate boiling heat transfer mechanism which predominates in the high heat flux region of the engine during high power level operation.

  16. Pressurized Hybrid Heat Pipe for Passive IN-Core Cooling System (PINCs) in Advanced Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-05-15

    The representative operating limit of the thermosyphon heat pipe is flooding limit that arises from the countercurrent flow of vapor and liquid. The effect of difference between wetted perimeter and heated perimeter on the flooding limit of the thermosyphons has not been studied; despite the effect of cross-sectional area of the vapor path on the heat transfer characteristics of thermosyphons have been studied. Additionally, the hybrid heat pipe must operate at the high temperature and high pressure environment because it will be inserted to the active core to remove the decay heat. However, the previously studied heat pipes operated below the atmospheric pressure. Therefore, the effect of the unique geometry for hybrid heat pipe and operating pressure on the heat transfer characteristics including the flooding limit of hybrid heat pipe was experimentally measured. Hybrid heat pipe as a new conceptual decay heat removal device was proposed. For the development of hybrid heat pipe operating at high temperature and high pressure conditions, the pressurized hybrid heat pipe was prepared and the thermal performances including operation limits of hybrid heat pipe were experimentally measured. Followings were obtained: (1) As operating pressure of the heat pipe increases, the evaporation heat transfer coefficient increases due to heat transfer with convective pool boiling mode. (2) Non-condensable gas charged in the test section for the pressurization lowered the condensation heat transfer by impeding the vapor flow to the condenser. (3) The deviations between experimentally measured flooding limits for hybrid heat pipes and the values from correlation for annular thermosyphon were observed.

  17. 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)

    Horacek, L.; Palyza, J.; Zdarek, J.; Vizina, M.

    2004-01-01

    Core Cooling System) of Temelin NPP. Review of results, conclusions and lessons learned obtained within the Pilot study are discussed and presented. (author)

  18. Cooling water injection system

    International Nuclear Information System (INIS)

    Inai, Nobuhiko.

    1989-01-01

    In a BWR type reactor, ECCS system is constituted as a so-called stand-by system which is not used during usual operation and there is a significant discontinuity in relation with the usual system. It is extremely important that ECCS operates upon occurrence of accidents just as specified. In view of the above in the present invention, the stand-by system is disposed along the same line with the usual system. That is, a driving water supply pump for supplying driving water to a jet pump is driven by a driving mechanism. The driving mechanism drives continuously the driving water supply pump in a case if an expected accident such as loss of the function of the water supply pump, as well as during normal operation. That is, all of the water supply pump, jet pump, driving water supply pump and driving mechanism therefor are caused to operate also during normal operation. The operation of them are not initiated upon accident. Thus, the cooling water injection system can perform at high reliability to remarkably improve the plant safety. (K.M.)

  19. Modelization of cooling system components

    Energy Technology Data Exchange (ETDEWEB)

    Copete, Monica; Ortega, Silvia; Vaquero, Jose Carlos; Cervantes, Eva [Westinghouse Electric (Spain)

    2010-07-01

    In the site evaluation study for licensing a new nuclear power facility, the criteria involved could be grouped in health and safety, environment, socio-economics, engineering and cost-related. These encompass different aspects such as geology, seismology, cooling system requirements, weather conditions, flooding, population, and so on. The selection of the cooling system is function of different parameters as the gross electrical output, energy consumption, available area for cooling system components, environmental conditions, water consumption, and others. Moreover, in recent years, extreme environmental conditions have been experienced and stringent water availability limits have affected water use permits. Therefore, modifications or alternatives of current cooling system designs and operation are required as well as analyses of the different possibilities of cooling systems to optimize energy production taking into account water consumption among other important variables. There are two basic cooling system configurations: - Once-through or Open-cycle; - Recirculating or Closed-cycle. In a once-through cooling system (or open-cycle), water from an external water sources passes through the steam cycle condenser and is then returned to the source at a higher temperature with some level of contaminants. To minimize the thermal impact to the water source, a cooling tower may be added in a once-through system to allow air cooling of the water (with associated losses on site due to evaporation) prior to returning the water to its source. This system has a high thermal efficiency, and its operating and capital costs are very low. So, from an economical point of view, the open-cycle is preferred to closed-cycle system, especially if there are no water limitations or environmental restrictions. In a recirculating system (or closed-cycle), cooling water exits the condenser, goes through a fixed heat sink, and is then returned to the condenser. This configuration

  20. AGN Heating in Simulated Cool-core Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuan; Ruszkowski, Mateusz [Department of Astronomy, University of Michigan, 1085 S. University Avenue, Ann Arbor, MI 48109 (United States); Bryan, Greg L., E-mail: yuanlium@umich.edu [Department of Astronomy, Columbia University, Pupin Physics Laboratories, New York, NY 10027 (United States)

    2017-10-01

    We analyze heating and cooling processes in an idealized simulation of a cool-core cluster, where momentum-driven AGN feedback balances radiative cooling in a time-averaged sense. We find that, on average, energy dissipation via shock waves is almost an order of magnitude higher than via turbulence. Most of the shock waves in the simulation are very weak shocks with Mach numbers smaller than 1.5, but the stronger shocks, although rare, dissipate energy more effectively. We find that shock dissipation is a steep function of radius, with most of the energy dissipated within 30 kpc, more spatially concentrated than radiative cooling loss. However, adiabatic processes and mixing (of post-shock materials and the surrounding gas) are able to redistribute the heat throughout the core. A considerable fraction of the AGN energy also escapes the core region. The cluster goes through cycles of AGN outbursts accompanied by periods of enhanced precipitation and star formation, over gigayear timescales. The cluster core is under-heated at the end of each cycle, but over-heated at the peak of the AGN outburst. During the heating-dominant phase, turbulent dissipation alone is often able to balance radiative cooling at every radius but, when this is occurs, shock waves inevitably dissipate even more energy. Our simulation explains why some clusters, such as Abell 2029, are cooling dominated, while in some other clusters, such as Perseus, various heating mechanisms including shock heating, turbulent dissipation and bubble mixing can all individually balance cooling, and together, over-heat the core.

  1. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

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

    1983-03-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

    for extended operational life of the AGR fleet. To this aim specimens taken from actual cores of the leading stations as well as virgin material will be irradiated in both oxidising and inert environment. It requires the design and operation of a carbon dioxide purge and gas handling system for the irradiation stages, allowing on-line gas composition monitoring and control. Furthermore the high weight losses anticipated require a targeted verification of the validity range of existing test methods, and potentially their extension, or even development of new tests. The presentation is an overview of the irradiation and post-irradiation testing projects for present and future gas cooled reactors, and will highlight results and expectancies. (authors)

  3. Isolated core vs. superficial cooling effects on virtual maze navigation.

    Science.gov (United States)

    Payne, Jennifer; Cheung, Stephen S

    2007-07-01

    Cold impairs cognitive performance and is a common occurrence in many survival situations. Altered behavior patterns due to impaired navigation abilities in cold environments are potential problems in lost-person situations. We investigated the separate effects of low core temperature and superficial cooling on a spatially demanding virtual navigation task. There were 12 healthy men who were passively cooled via 15 degrees C water immersion to a core temperature of 36.0 degrees C, then transferred to a warm (40 degrees C) water bath to eliminate superficial shivering while completing a series of 20 virtual computer mazes. In a control condition, subjects rested in a thermoneutral (approximately 35 degrees C) bath for a time-matched period before being transferred to a warm bath for testing. Superficial cooling and distraction were achieved by whole-body immersion in 35 degree water for a time-matched period, followed by lower leg immersion in 10 degree C water for the duration of the navigational tests. Mean completion time and mean error scores for the mazes were not significantly different (p > 0.05) across the core cooling (16.59 +/- 11.54 s, 0.91 +/- 1.86 errors), control (15.40 +/- 8.85 s, 0.82 +/- 1.76 errors), and superficial cooling (15.19 +/- 7.80 s, 0.77 +/- 1.40 errors) conditions. Separately reducing core temperature or increasing cold sensation in the lower extremities did not influence performance on virtual computer mazes, suggesting that navigation is more resistive to cooling than other, simpler cognitive tasks. Further research is warranted to explore navigational ability at progressively lower core and skin temperatures, and in different populations.

  4. Thermohydraulics of emergency core cooling in light water reactors

    International Nuclear Information System (INIS)

    1989-10-01

    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

  5. Post-implementation review of inadequate core cooling instrumentation

    International Nuclear Information System (INIS)

    Anderson, J.L.; Anderson, R.L.; Hagen, E.W.; Morelock, T.C.; Huang, T.L.; Phillips, L.E.

    1988-01-01

    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)

  6. COOLING STAGES OF CRYOGENIC SYSTEMS

    OpenAIRE

    Троценко, А. В.

    2011-01-01

    The formalized definition for cooling stage of low temperature system is done. Based on existing information about the known cryogenic unit cycles the possible types of cooling stages are single out. From analyses of these stages their classification by various characteristics is suggested. The results of thermodynamic optimization of final throttle stage of cooling, which are used as working fluids helium, hydrogen and nitrogen, are shown.

  7. 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.

  8. 46 CFR 153.432 - Cooling systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cooling systems. 153.432 Section 153.432 Shipping COAST... Control Systems § 153.432 Cooling systems. (a) Each cargo cooling system must have an equivalent standby... cooling system. (b) Each tankship that has a cargo tank with a required cooling system must have a manual...

  9. Heat removal capability of core-catcher with inclined cooling channels

    International Nuclear Information System (INIS)

    Suzuki, Y.; Tahara, M.; Kurita, T.; Hamazaki, R.; Morooka, S.

    2009-01-01

    A core-catcher is one of the mitigation systems that provide functions of molten corium cooling and stabilization during a severe accident. Toshiba has been developing a compact core-catcher to be placed at the lower drywell floor in the containment vessel for the next generation BWR as well as near term ABWR. This paper presents the evaluation of heat removal capability of the core-catcher with inclined cooling channels, our verification status and plan. The heat removal capability of the core-catcher is analyzed by using the newly developed two-phase flow analysis code which incorporates drift flux parameters for inclined channels and the CHF correlation obtained from SULTAN tests. Effects of geometrical parameters such as the inclination and the gap size of the cooling channel on the heat removal capability are also evaluated. These results show that the core-catcher has sufficient capability to cool the molten corium during a severe accident. Based on the analysis, it has been shown that the core-catcher has an efficient capability of heat removal to cool the molten corium. (author)

  10. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

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

  11. Cooling water systems design using process integration

    CSIR Research Space (South Africa)

    Gololo, KV

    2010-09-01

    Full Text Available Cooling water systems are generally designed with a set of heat exchangers arranged in parallel. This arrangement results in higher cooling water flowrate and low cooling water return temperature thus reducing cooling tower efficiency. Previous...

  12. A core management system for JRR-3

    International Nuclear Information System (INIS)

    Soyama, Kazuhiko; Tsuruta, Harumichi; Ichikawa, Hiroki; Nemoto, Hiroyuki.

    1991-05-01

    Japan Research Reactor No.3 (JRR-3) was upgraded to the thermal output with 20 MW by replacing the core, cooling system and utilization facilities. It is a water moderated and cooled, pool type reactor using 20% enriched U · Alx fuel. A core management system for JRR-3 has been made. This code system can manage of reactivity, power distribution and burn up in consideration of the position of control rod, fuel arrangement and operation pattern. This report is the user's manual of this code system. (author)

  13. Superconducting magnet cooling system

    Science.gov (United States)

    Vander Arend, Peter C.; Fowler, William B.

    1977-01-01

    A device is provided for cooling a conductor to the superconducting state. The conductor is positioned within an inner conduit through which is flowing a supercooled liquid coolant in physical contact with the conductor. The inner conduit is positioned within an outer conduit so that an annular open space is formed therebetween. Through the annular space is flowing coolant in the boiling liquid state. Heat generated by the conductor is transferred by convection within the supercooled liquid coolant to the inner wall of the inner conduit and then is removed by the boiling liquid coolant, making the heat removal from the conductor relatively independent of conductor length.

  14. Cryogenic cooling system for HTS cable

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Shigeru [Taiyo Nippon Sanso, Tsukuba (Japan)

    2017-06-15

    Recently, Research and development activity of HTS (High Temperature Superconducting) power application is very progressive worldwide. Especially, HTS cable system and HTSFCL (HTS Fault current limiter) system are proceeding to practical stages. In such system and equipment, cryogenic cooling system, which makes HTS equipment cooled lower than critical temperature, is one of crucial components. In this article, cryogenic cooling system for HTS application, mainly cable, is reviewed. Cryogenic cooling system can be categorized into conduction cooling system and immersion cooling system. In practical HTS power application area, immersion cooling system with sub-cooled liquid nitrogen is preferred. The immersion cooling system is besides grouped into open cycle system and closed cycle system. Turbo-Brayton refrigerator is a key component for closed cycle system. Those two cooling systems are focused in this article. And, each design and component of the cooling system is explained.

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

    International Nuclear Information System (INIS)

    Jelinek, Tomas

    2015-01-01

    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

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

    International Nuclear Information System (INIS)

    2013-11-01

    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

  17. Core melt retention and cooling concept of the ERP

    Energy Technology Data Exchange (ETDEWEB)

    Weisshaeupl, H [SIEMENS/KWU, Erlangen (Germany); Yvon, M [Nuclear Power International, Paris (France)

    1996-12-01

    For the French/German European Pressurized Water Reactor (EPR) mitigative measures to cope with the event of a severe accident with core melt down are considered already at the design stage. Following the course of a postulated severe accident with reactor pressure vessel melt through one of the most important features of a future design must be to stabilize and cool the melt within the containment by dedicated measures. This measures should - as far as possible - be passive. One very promising solution for core melt retention seems to be a large enough spreading of the melt on a high temperature resistant protection layer with water cooling from above. This is the favorite concept for the EPR. In dealing with the retention of a molten core outside of the RPV several ``steps`` from leaving the RPV to finally stabilize the melt have to gone through. These steps are: collection of the melt; transfer of the melt; distribution of the melt; confining; cooling and stabilization. The technical features for the EPR solution of a large spreading of the melt are: Dedicated spreading chamber outside the reactor pit (area about 150 m{sup 2}); high temperature resistant protection layers (e.g. Zirconia bricks) at the bottom and part of the lateral structures (thus avoiding melt concrete interaction); reactor pit and spreading compartment are connected via a discharge channel which has a slope to the spreading area and is closed by a steel plate, which will resist the core melt for a certain time in order to allow a collection of the melt; the spreading compartments is connected with the In-Containment Refuelling Water Storage Tank (IRWST) with pipes for water flooding after spreading. These pipes are closed and will only be opened by the hot melt itself. It is shown how the course of the different steps mentioned above is processed and how each of these steps is automatically and passively achieved. (Abstract Truncated)

  18. Modelling aerosol behavior in reactor cooling systems

    International Nuclear Information System (INIS)

    McDonald, B.H.

    1990-01-01

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

  19. Proceedings: Cooling tower and advanced cooling systems conference

    International Nuclear Information System (INIS)

    1995-02-01

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

  20. Forced draft wet cooling systems

    International Nuclear Information System (INIS)

    Daubert, A.; Caudron, L.; Viollet, P.L.

    1975-01-01

    The disposal of the heat released from a 1000MW power plant needs a natural draft tower of about 130m of diameter at the base, and 170m height, or a cooling system with a draft forced by about forty vans, a hundred meters in diameter, and thirty meters height. The plumes from atmospheric cooling systems form, in terms of fluid mechanics, hot jets in a cross current. They consist in complex flows that must be finely investigated with experimental and computer means. The study, currently being performed at the National Hydraulics Laboratory, shows that as far as the length and height of visible plumes are concerned, the comparison is favorable to some types of forced draft cooling system, for low and medium velocities, (below 5 or 6m/s at 10m height. Beyond these velocities, the forced draft sends the plume up to smaller heights, but the plume is generally more dilute [fr

  1. TPX heating and cooling system

    International Nuclear Information System (INIS)

    Kungl, D.J.; Knutson, D.S.; Costello, J.; Stoenescu, S.; Yemin, L.

    1995-01-01

    TPX, while having primarily super-conducting coils that do not require water cooling, still has very significant water cooling requirements for the plasma heating systems, vacuum vessel, plasma facing components, diagnostics, and ancillary equipment. This is accentuated by the 1000-second pulse requirement. Two major design changes, which have significantly affected the TPX Heating and Cooling System, have been made since the conceptual design review in March of 1993. This paper will discuss these changes and review the current status of the conceptual design. The first change involves replacing the vacuum vessel neutron shielding configuration of lead/glass composite tile by a much simpler and more reliable borated water shield. The second change reduces the operating temperature of the vacuum vessel from 150 C to ≥50 C. With this temperature reduction, all in-vessel components and the vessel will be supplied by coolant at a common ≥50 C inlet temperature. In all, six different heating and cooling supply requirements (temperature, pressure, water quality) for the various TPX components must be met. This paper will detail these requirements and provide an overview of the Heating and Cooling System design while focusing on the ramifications of the TPX changes described above

  2. Core design of a high breeding fast reactor cooled by supercritical pressure light water

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Takayuki, E-mail: russell@ruri.waseda.jp; Yamaji, Akifumi

    2016-01-15

    Highlights: • Core design concept of supercritical light water cooled fast breeding reactor is developed. • Compound system doubling time (CSDT) is applied for considering an appropriate target of breeding performance. • Breeding performance is improved by reducing fuel rod diameter of the seed assembly. • Core pressure loss is reduced by enlarging the coolant channel area of the seed assembly. - Abstract: A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

  3. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Uspuras, E.; Kaliatka, A.

    2001-01-01

    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)

  5. Core design studies on various forms of coolants and fuel materials. 2. Studies on liquid heavy metal and gas cooled cores, small cores and evaluation of 4-type cores

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Sakashita, Yoshiyuki; Naganuma, Masayuki; Takaki, Naoyuki; Mizuno, Tomoyasu; Ikegami, Tetsuo

    2001-01-01

    Alternative concepts to sodium cooled fast reactors, such as heavy metal liquid cooled reactors and gas cooled fast reactors were studied in Phase-1 of the feasibility studies, aiming at simplification of the system, high thermal efficiency and enhancing safety. Fuel and core specifications and nuclear characteristics were surveyed to meet the targets for commercialization of fast reactor cycle. Nuclear characteristics of small fast reactor cores were also surveyed from the perspective of the possibility of multi-purpose use and dispersed power stations. The key points of the design study for each concept in Phase-2 were summarized from the aspect of the screening of the candidates for FR commercialization. (author)

  6. Testing the Large-scale Environments of Cool-core and Non-cool-core Clusters with Clustering Bias

    Energy Technology Data Exchange (ETDEWEB)

    Medezinski, Elinor; Battaglia, Nicholas; Cen, Renyue; Gaspari, Massimo; Strauss, Michael A.; Spergel, David N. [Department of Astrophysical Sciences, 4 Ivy Lane, Princeton, NJ 08544 (United States); Coupon, Jean, E-mail: elinorm@astro.princeton.edu [Department of Astronomy, University of Geneva, ch. dEcogia 16, CH-1290 Versoix (Switzerland)

    2017-02-10

    There are well-observed differences between cool-core (CC) and non-cool-core (NCC) clusters, but the origin of this distinction is still largely unknown. Competing theories can be divided into internal (inside-out), in which internal physical processes transform or maintain the NCC phase, and external (outside-in), in which the cluster type is determined by its initial conditions, which in turn leads to different formation histories (i.e., assembly bias). We propose a new method that uses the relative assembly bias of CC to NCC clusters, as determined via the two-point cluster-galaxy cross-correlation function (CCF), to test whether formation history plays a role in determining their nature. We apply our method to 48 ACCEPT clusters, which have well resolved central entropies, and cross-correlate with the SDSS-III/BOSS LOWZ galaxy catalog. We find that the relative bias of NCC over CC clusters is b = 1.42 ± 0.35 (1.6 σ different from unity). Our measurement is limited by the small number of clusters with core entropy information within the BOSS footprint, 14 CC and 34 NCC clusters. Future compilations of X-ray cluster samples, combined with deep all-sky redshift surveys, will be able to better constrain the relative assembly bias of CC and NCC clusters and determine the origin of the bimodality.

  7. Testing the Large-scale Environments of Cool-core and Non-cool-core Clusters with Clustering Bias

    International Nuclear Information System (INIS)

    Medezinski, Elinor; Battaglia, Nicholas; Cen, Renyue; Gaspari, Massimo; Strauss, Michael A.; Spergel, David N.; Coupon, Jean

    2017-01-01

    There are well-observed differences between cool-core (CC) and non-cool-core (NCC) clusters, but the origin of this distinction is still largely unknown. Competing theories can be divided into internal (inside-out), in which internal physical processes transform or maintain the NCC phase, and external (outside-in), in which the cluster type is determined by its initial conditions, which in turn leads to different formation histories (i.e., assembly bias). We propose a new method that uses the relative assembly bias of CC to NCC clusters, as determined via the two-point cluster-galaxy cross-correlation function (CCF), to test whether formation history plays a role in determining their nature. We apply our method to 48 ACCEPT clusters, which have well resolved central entropies, and cross-correlate with the SDSS-III/BOSS LOWZ galaxy catalog. We find that the relative bias of NCC over CC clusters is b = 1.42 ± 0.35 (1.6 σ different from unity). Our measurement is limited by the small number of clusters with core entropy information within the BOSS footprint, 14 CC and 34 NCC clusters. Future compilations of X-ray cluster samples, combined with deep all-sky redshift surveys, will be able to better constrain the relative assembly bias of CC and NCC clusters and determine the origin of the bimodality.

  8. System for cooling a cabinet

    DEFF Research Database (Denmark)

    2015-01-01

    The present disclosure relates to a cooling system comprising an active magnetic regenerator having a cold side and a hot side, a hot side heat exchanger connected to the hot side of the magnetic regenerator, one or more cold side heat exchangers, and a cold store reservoir comprising a volume...

  9. Lamination cooling system formation method

    Science.gov (United States)

    Rippel, Wally E [Altadena, CA; Kobayashi, Daryl M [Monrovia, CA

    2009-05-12

    An electric motor, transformer or inductor having a cooling system. A stack of laminations have apertures at least partially coincident with apertures of adjacent laminations. The apertures define straight or angled cooling-fluid passageways through the lamination stack. Gaps between the adjacent laminations are sealed by injecting a heat-cured sealant into the passageways, expelling excess sealant, and heat-curing the lamination stack. Manifold members adjoin opposite ends of the lamination stack, and each is configured with one or more cavities to act as a manifold to adjacent passageway ends. Complex manifold arrangements can create bidirectional flow in a variety of patterns.

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

    Science.gov (United States)

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

    2018-02-01

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

  11. Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

    Science.gov (United States)

    Darmawan, R.

    2018-01-01

    Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

  12. Gas hydrate cool storage system

    Science.gov (United States)

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

    1984-09-12

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

  13. Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Su Xia

    2013-01-01

    The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

  14. Fluid dynamic simulation of the non homogeneous steam-air mixture motion in the dome of MARS safety core cooling system

    International Nuclear Information System (INIS)

    Said, S.A.; Caira, M.; Gramiccia, L.; Naviglio, A.

    1992-01-01

    One of the main features of the MARS, an inherently safe nuclear reactor of the new generation, is the innovative decay heat removal system. This has a high inherent reliability thanks to the complete absence of active components. The core decay heat is removed by the vaporization of the water in an emergency reservoir; then the steam collected in the dome over the pool condenses in the air condenser and returns back to the reservoir creating a heat sink of nearly infinite capacity. The transient fluid dynamic numerical simulation of the steam-air mixture flow in the dome is presented. This allows an assessment to be made of the time required for the uncondensable gases to be evacuated. After that time the condenser works at its rated capacity. (4 figures) (Author)

  15. A comparative design study of PB-BI cooled reactor cores with forced and natural convection cooling

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Enuma, Yasuhiro; Tanji, Mikio

    2003-01-01

    A comparative core design study is performed on Pb-Bi cooled reactors with forced and natural convection (FC and NC) cooling. Major interests of the study are core performance and core safety features. The designed core concepts with nitride fuel achieve reasonable breeding capability. The results of unprotected event analyses such as UTOP and ULOF show that both of concepts have possible features to withstand unprotected events due to negative reactivity feedback by Doppler effect, control rod drive line expansion, etc. These results lead to a conclusion that both of concepts have possible capability as one of future promising core concepts. A FC cooling core concept has more advantage if fuel recycle viewpoint is emphasized. (author)

  16. Cooling Tower Overhaul of Secondary Cooling System in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Park, Young Chul; Lee, Young Sub; Jung, Hoan Sung; Lim, In Chul [KAERI, Daejeon (Korea, Republic of)

    2007-07-01

    HANARO, an open-tank-in-pool type research reactor of 30 MWth power in Korea, has been operating normally since its initial criticality in February, 1995. For the last about ten years, A cooling tower of a secondary cooling system has been operated normally in HANARO. Last year, the cooling tower has been overhauled for preservative maintenance including fills, eliminators, wood support, water distribution system, motors, driving shafts, gear reducers, basements, blades and etc. This paper describes the results of the overhaul. As results, it is confirmed that the cooling tower maintains a good operability through a filed test. And a cooling capability will be tested when a wet bulb temperature is maintained about 28 .deg. C in summer and the reactor is operated with the full power.

  17. Experiment of IEA-R1 reactor core cooling by air convection after pool water loss accident

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Baptista Filho, Benedito Dias

    2000-01-01

    This paper presents a study of a Emergency Core Cooling to be applied to the IEA-R1 reactor. This system must have the characteristics of passive action, with water spraying over the core, and feeding by gravity from elevated reservoirs. In the evaluation, this system must demonstrate that when the reservoirs are emptied, the core cooling must assure to be fulfilled by air natural convection. This work presents the results of temperature distribution in a test section with plates electrically heated simulation the heat generation conditions on the most heated reactor element

  18. Information technology equipment cooling system

    Science.gov (United States)

    Schultz, Mark D.

    2014-06-10

    According to one embodiment, a system for removing heat from a rack of information technology equipment may include a sidecar indoor air to liquid heat exchanger that cools warm air generated by the rack of information technology equipment. The system may also include a liquid to liquid heat exchanger and an outdoor heat exchanger. The system may further include configurable pathways to connect and control fluid flow through the sidecar heat exchanger, the liquid to liquid heat exchanger, the rack of information technology equipment, and the outdoor heat exchanger based upon ambient temperature and/or ambient humidity to remove heat from the rack of information technology equipment.

  19. Cooling system for superconducting magnet

    Science.gov (United States)

    Gamble, Bruce B.; Sidi-Yekhlef, Ahmed

    1998-01-01

    A cooling system is configured to control the flow of a refrigerant by controlling the rate at which the refrigerant is heated, thereby providing an efficient and reliable approach to cooling a load (e.g., magnets, rotors). The cooling system includes a conduit circuit connected to the load and within which a refrigerant circulates; a heat exchanger, connected within the conduit circuit and disposed remotely from the load; a first and a second reservoir, each connected within the conduit, each holding at least a portion of the refrigerant; a heater configured to independently heat the first and second reservoirs. In a first mode, the heater heats the first reservoir, thereby causing the refrigerant to flow from the first reservoir through the load and heat exchanger, via the conduit circuit and into the second reservoir. In a second mode, the heater heats the second reservoir to cause the refrigerant to flow from the second reservoir through the load and heat exchanger via the conduit circuit and into the first reservoir.

  20. Preliminary design of a borax internal core-catcher for a gas cooled fast reactor

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Dorner, S.; Schumacher, G.

    1976-09-01

    Preliminary thermal calculations show that a core-catcher appears to be feasible, which is able to cope with the complete meltdown of the core and blankets of a 1,000 MWe GCFR. This core-catcher is based on borax (Na 2 B 4 O 7 ) as dissolving material of the oxide fuel and of the fission products occuring in oxide form. The borax is contained in steel boxes forming a 2.1 meter thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel, just underneath the reactor core. The fission products are dispersed in the pool formed by the liquid borax. The heat power density in the pool is conveniently reduced and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system. (orig.) [de

  1. 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.

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

    International Nuclear Information System (INIS)

    Grasso, G.; Petrovich, C.; Mattioli, D.; Artioli, C.; Sciora, P.; Gugiu, D.; Bandini, G.; Bubelis, E.; Mikityuk, K.

    2014-01-01

    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 th ) and of its demonstrator reactor (300 MW 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

  3. 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.

  4. Evaluation of long-term post-accident core cooling of Three Mile Island Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-04-15

    On the basis of current understanding of the accident scenario and available data, the staff reports here on its evaluation of the condition of the core and the core flow resistance as it might affect ability to cool the core by natural circulation. The natural circulation cooling capability of TMI-2 for the estimated core flow resistance and a variety of other conditions is evaluated and a comparison of the Base Case and off-nominal plant configurations is presented. The potential for and effects of natural convection core cooling are addressed, and the staff recommendations for reactor performance acceptance criteria upon initiation of natural convection are presented.

  5. Turbine airfoil with ambient cooling system

    Science.gov (United States)

    Campbell, Jr, Christian X.; Marra, John J.; Marsh, Jan H.

    2016-06-07

    A turbine airfoil usable in a turbine engine and having at least one ambient air cooling system is disclosed. At least a portion of the cooling system may include one or more cooling channels configured to receive ambient air at about atmospheric pressure. The ambient air cooling system may have a tip static pressure to ambient pressure ratio of at least 0.5, and in at least one embodiment, may include a tip static pressure to ambient pressure ratio of between about 0.5 and about 3.0. The cooling system may also be configured such that an under root slot chamber in the root is large to minimize supply air velocity. One or more cooling channels of the ambient air cooling system may terminate at an outlet at the tip such that the outlet is aligned with inner surfaces forming the at least one cooling channel in the airfoil to facilitate high mass flow.

  6. Simulations of floor cooling system capacity

    International Nuclear Information System (INIS)

    Odyjas, Andrzej; Górka, Andrzej

    2013-01-01

    Floor cooling system capacity depends on its physical and operative parameters. Using numerical simulations, it appears that cooling capacity of the system largely depends on the type of cooling loads occurring in the room. In the case of convective cooling loads capacity of the system is small. However, when radiation flux falls directly on the floor the system significantly increases productivity. The article describes the results of numerical simulations which allow to determine system capacity in steady thermal conditions, depending on the type of physical parameters of the system and the type of cooling load occurring in the room. Moreover, the paper sets out the limits of system capacity while maintaining a minimum temperature of the floor surface equal to 20 °C. The results are helpful for designing system capacity in different type of cooling loads and show maximum system capacity in acceptable thermal comfort condition. -- Highlights: ► We have developed numerical model for simulation of floor cooling system. ► We have described floor system capacity depending on its physical parameters. ► We have described floor system capacity depending on type of cooling loads. ► The most important in the obtained cooling capacities is the type of cooling loads. ► The paper sets out the possible maximum cooling floor system capacity

  7. Effects of running time of a cattle-cooling system on core body temperature of cows on dairy farms in an arid environment.

    Science.gov (United States)

    Ortiz, X A; Smith, J F; Bradford, B J; Harner, J P; Oddy, A

    2010-10-01

    Two experiments were conducted on a commercial dairy farm to describe the effects of a reduction in Korral Kool (KK; Korral Kool Inc., Mesa, AZ) system operating time on core body temperature (CBT) of primiparous and multiparous cows. In the first experiment, KK systems were operated for 18, 21, or 24 h/d while CBT of 63 multiparous Holstein dairy cows was monitored. All treatments started at 0600 h, and KK systems were turned off at 0000 h and 0300 h for the 18-h and 21-h treatments, respectively. Animals were housed in 9 pens and assigned randomly to treatment sequences in a 3 × 3 Latin square design. In the second experiment, 21 multiparous and 21 primiparous cows were housed in 6 pens and assigned randomly to treatment sequences (KK operated for 21 or 24 h/d) in a switchback design. All treatments started at 0600 h, and KK systems were turned off at 0300 h for the 21-h treatments. In experiment 1, cows in the 24-h treatment had a lower mean CBT than cows in the 18- and 21-h treatments (38.97, 39.08, and 39.03±0.04°C, respectively). The significant treatment by time interaction showed that the greatest treatment effects occurred at 0600 h; treatment means at this time were 39.43, 39.37, and 38.88±0.18°C for 18-, 21-, and 24-h treatments, respectively. These results demonstrate that a reduction in KK system running time of ≥3 h/d will increase CBT. In experiment 2, a significant parity by treatment interaction was found. Multiparous cows on the 24-h treatment had lower mean CBT than cows on the 21-h treatment (39.23 and 39.45±0.17°C, respectively), but treatment had no effect on mean CBT of primiparous cows (39.50 and 39.63±0.20°C for 21- and 24-h treatments, respectively). A significant treatment by time interaction was observed, with the greatest treatment effects occurring at 0500 h; treatment means at this time were 39.57, 39.23, 39.89, and 39.04±0.24°C for 21-h primiparous, 24-h primiparous, 21-h multiparous, and 24-h multiparous cows

  8. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Seon Oh; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

    2017-08-15

    The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  9. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Directory of Open Access Journals (Sweden)

    Seon Oh Yu

    2017-08-01

    Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  10. Reactor core of light water-cooled reactor

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Lakkis, I.A.

    1993-01-01

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

  12. Analysis of emergency core cooling capability of direct vessel vertical injection using CFX

    International Nuclear Information System (INIS)

    Yoon, Sang H.; Yu, Yong H.; Suh, Kune Y.

    2003-01-01

    More reliable and efficient safety injection system is of utmost importance in the design of advanced reactors such as the APR1400 (Advanced Power Reactor 1400 MWe). In this work, a new idea is proposed to inject the Emergency Core Cooling (ECC) water utilizing a dedicated nozzle with a vertically downward elbow. The Direct Vessel Injection (DVI) system is located horizontally above the cold leg in the APR1400. However, the horizontal injection method may not always satisfy the ECC penetration requirement into the core on account of rather involved multidimensional thermal and hydraulic phenomena occurring in the annular reactor downcomer such as bypass, impingement, entrainment and sweepout, condensation oscillation, etc. Thus, a novel concept is called for from the reactor safety point of view. The Direct Vessel Vertical Injection (DVVI) system is one of these efforts to penetrate as much the ECC water through the downcomer into the core as is practically achievable. The DVVI system can increase the momentum of the downward flow, thus minimizing the effect of water impingement on the core barrel and the direct bypass though the break. To support the claim of increased downward momentum of flow in the DVVI system, computational fluid dynamics analyses were performed using CFX. The new concept of the DVVI system, which can certainly help increase the core thermal margin, is found to be more efficient than DVI. If the structural problem in the manufacturing process is properly solved, this concept can safely be applied in the advanced nuclear reactor design

  13. Environmental effects of cooling systems

    International Nuclear Information System (INIS)

    1980-01-01

    Since the International Atomic Energy Agency published in 1974 Thermal Discharges at Nuclear Power Stations (Technical Reports Series No.155), much progress has been made in the understanding of phenomena related to thermal discharges. Many studies have been performed in Member States and from 1973 to 1978 the IAEA sponsored a co-ordinated research programme on 'Physical and Biological Effects on the Environment of Cooling Systems and Thermal Discharges from Nuclear Power Stations'. Seven laboratories from Canada, the Federal Republic of Germany, India and the United States of America were involved in this programme, and a lot of new information has been obtained during the five years' collaboration. The progress of the work was discussed at annual co-ordination meetings and the results are presented in the present report. It complements the previous report mentioned above as it deals with several questions that were not answered in 1974. With the conclusion of this co-ordinated programme, it is obvious that some problems have not yet been resolved and that more work is necessary to assess completely the impact of cooling systems on the environment. It is felt, however, that the data gathered here will bring a substantial contribution to the understanding of the subject

  14. Evaporative cooling enhanced cold storage system

    Science.gov (United States)

    Carr, P.

    1991-10-15

    The invention provides an evaporatively enhanced cold storage system wherein a warm air stream is cooled and the cooled air stream is thereafter passed into contact with a cold storage unit. Moisture is added to the cooled air stream prior to or during contact of the cooled air stream with the cold storage unit to effect enhanced cooling of the cold storage unit due to evaporation of all or a portion of the added moisture. Preferably at least a portion of the added moisture comprises water condensed during the cooling of the warm air stream. 3 figures.

  15. Solar-powered cooling system

    Science.gov (United States)

    Farmer, Joseph C.

    2015-07-28

    A solar-powered adsorption-desorption refrigeration and air conditioning system that uses nanostructural materials such as aerogels, zeolites, and sol gels as the adsorptive media. Refrigerant molecules are adsorbed on the high surface area of the nanostructural material while the material is at a relatively low temperature, perhaps at night. During daylight hours, when the nanostructural materials is heated by the sun, the refrigerant are thermally desorbed from the surface of the aerogel, thereby creating a pressurized gas phase in the vessel that contains the aerogel. This solar-driven pressurization forces the heated gaseous refrigerant through a condenser, followed by an expansion valve. In the condenser, heat is removed from the refrigerant, first by circulating air or water. Eventually, the cooled gaseous refrigerant expands isenthalpically through a throttle valve into an evaporator, in a fashion similar to that in more conventional vapor recompression systems.

  16. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

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

    International Nuclear Information System (INIS)

    Rouge, S.; Seiler, J.M.

    1994-09-01

    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/m 2 ) 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/m 2 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

  18. Thermal-hydraulic analysis of the OSURR pool for power upgrade with natural convection core cooling

    International Nuclear Information System (INIS)

    Ha, J.J.; Aldemir, T.

    1988-01-01

    Natural convection mode core cooling will be maintained in the LEU conversion/power upgrade of The Ohio State University Research Reactor (OSURR) to 250-500 kW. The pool water will be cooled by a water-glycol-air and a water-water heat exchanger. A plume disperser will be installed in the pool to minimize evaporation from the pool top and to maintain the dose rate due to N-16 activity within allowable levels. The minimization of the pool heat removal system operation costs necessitates maximizing the inlet temperature to the water-glycol-air heat exchanger. For the maximization process, the change in the pool temperature and velocity fields have to be investigated as a function of: location and orientation of the heat removal system components and the plume disperser in the pool; mass flow rate through the plume disperser. The velocity and temperature fields in the pool are determined using COMMIX-1A. The computational system model accounts for the presence of all the pool components (i.e. core, thermal column, beam ports, ion chamber, guide tubes, rabbit, neutron source etc.). The results show that: (1) Both the heat removal system inlet point and the plume disperser have to be located close to the top of the core. (2) Using a disperser system consisting of several pipes may be more feasible than a single unit. (3) For high disperser flow, the disperser jet has to be almost parallel to the top of the core to prevent flow reversal in coolant channels. (4) More than one disperser system may be necessary to create an inversion layer in the pool

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

    Science.gov (United States)

    McDonald, M.; Bayliss, M.; Benson, B. A.; Foley, R. J.; Ruel, J.; Sullivan, P.; Veilleux, S.; Aird, K. A.; Ashby, M. L. N.; Bautz, M.; hide

    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(sub 2-10 keV) = 8.2 10(exp 45) erg/s) galaxy cluster which hosts an extremely strong cooling flow (M(sub cool) = 3820 +/- 530 Stellar Mass/yr). Further, the central galaxy in this cluster appears to be experiencing a massive starburst (740 +/- 160 Stellar Mass/ yr), which suggests that the feedback source responsible for preventing runaway cooling in nearby cool core clusters may not yet be fully established in SPT-CLJ2344-4243. This large star formation rate implies that a significant fraction of the stars in the central galaxy of this cluster may form via accretion of the intracluster medium, rather than the current picture of central galaxies assembling entirely via mergers.

  20. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  1. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  2. Dry and mixed air cooling systems

    International Nuclear Information System (INIS)

    Gutner, Gidali.

    1975-01-01

    The various dry air cooling systems now in use or being developed are classified. The main dimensioning parameters are specified and the main systems already built are given with their characteristics. The available data allow dry air cooling to be situated against the other cooling modes and so specify the aim of the research or currently developed works. Some systems at development stages are briefly described. The interest in mixed cooling (assisted draft) and the principal available systems is analyzed. A program of research is outlined [fr

  3. Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

    2003-01-01

    The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

  4. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

    2017-07-11

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

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

    International Nuclear Information System (INIS)

    Kiefhaber, E.

    1978-10-01

    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

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

    Science.gov (United States)

    Johnson, Sean

    2017-08-01

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

  7. Core status computing system

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki.

    1982-01-01

    Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

  8. Neutronic design for a 100MWth Small modular natural circulation lead or lead-alloy cooled fast reactors core

    International Nuclear Information System (INIS)

    Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q.

    2015-01-01

    Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW th natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

  9. A combined capillary cooling system for cooling fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Ana Paula; Pelizza, Pablo Rodrigo; Galante, Renan Manozzo; Bazzo, Edson [Universidade Federal de Santa Catarina (LabCET/UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. Lab. de Combustao e Engenharia de Sistemas Termicos], Emails: ana@labcet.ufsc.br, pablo@labcet.ufsc.br, renan@labcet.ufsc.br, ebazzo@emc.ufsc.br

    2010-07-01

    The operation temperature control has an important influence over the PEMFC (Proton Exchange Membrane Fuel Cell) performance. A two-phase heat transfer system is proposed as an alternative for cooling and thermal control of PEMFC. The proposed system consists of a CPL (Capillary Pumped Loop) connected to a set of constant conductance heat pipes. In this work ceramic wick and stainless mesh wicks have been used as capillary structure of the CPL and heat pipes, respectively. Acetone has been used as the working fluid for CPL and deionized water for the heat pipes. Experimental results of three 1/4 inch stainless steel outlet diameter heats pipes and one CPL have been carried out and presented in this paper. Further experiments are planned coupling the proposed cooling system to a module which simulates the fuel cell. (author)

  10. Developments in power plant cooling systems

    International Nuclear Information System (INIS)

    Agarwal, N.K.

    1993-01-01

    A number of cooling systems are used in the power plants. The condenser cooling water system is one of the most important cooling systems in the plant. The system comprises a number of equipment. Plants using sea water for cooling are designed for the very high corrosion effects due to sea water. Developments are taking place in the design, materials of construction as well as protection philosophies for the various equipment. Power optimisation of the cycle needs to be done in order to design an economical system. Environmental (Protection) Act places certain limitations on the effluents from the plant. An attempt has been made in this paper to outline the developing trends in the various equipment in the condenser cooling water systems used at the inland as well as coastal locations. (author). 5 refs., 6 refs

  11. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  12. Conduction cooling systems for linear accelerator cavities

    Science.gov (United States)

    Kephart, Robert

    2017-05-02

    A conduction cooling system for linear accelerator cavities. The system conducts heat from the cavities to a refrigeration unit using at least one cavity cooler interconnected with a cooling connector. The cavity cooler and cooling connector are both made from solid material having a very high thermal conductivity of approximately 1.times.10.sup.4 W m.sup.-1 K.sup.-1 at temperatures of approximately 4 degrees K. This allows for very simple and effective conduction of waste heat from the linear accelerator cavities to the cavity cooler, along the cooling connector, and thence to the refrigeration unit.

  13. A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

    International Nuclear Information System (INIS)

    Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

    2002-01-01

    A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

  14. Hot gas path component cooling system

    Science.gov (United States)

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

    2014-02-18

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

  15. Conceptual core designs for a 1200 MWe sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

    2008-01-01

    The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

  16. Natural convection cooling of LEU cores for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Akhtar, K.M.

    1991-08-01

    The first high power and equilibrium LEU cores of PARR-1 have been analysed to assess the maximum operating power based on natural convection cooling, need for forced cooling to remove the decay heat and to estimate safety margins that commensurate with the predetermined power limit. Computer code NATCON and standard correlations have been used for the analysis. The parameters studied includes coolant velocity, temperature distribution in the core, heat fluxes at onset of nucleate boiling, pulsed boiling and burnup. (author)

  17. Spitzer mid-infrared spectra of cool-core galaxy clusters

    NARCIS (Netherlands)

    de Messières, G.E.; O'Connell, R.W.; McNamara, B.R.; Donahue, M.; Nulsen, P.E.J.; Voit, G.M.; Wise, M.W.; Smith, B.; Higdon, J.; Higdon, S.; Bastian, N.

    2010-01-01

    We have obtained mid-infrared spectra of nine cool-core galaxy clusters with the Infrared Spectrograph aboard the Spitzer Space Telescope. X-ray, ultraviolet and optical observations have demonstrated that each of these clusters hosts a cooling flow which seems to be fueling vigorous star formation

  18. The development of air cooled condensation systems

    International Nuclear Information System (INIS)

    Bodas, J.

    1990-01-01

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

  19. Evaluation of the gravity-injection capability for core cooling after a loss-of-SDC event

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Bang, Young Seok; Kim, Hho Jung

    1999-01-01

    In order to evaluate the gravity-drain capability to maintain core cooling after a loss-of-shutdown-cooling event during shutdown operation, the plant conditions of the Young Gwang Units 3 and 4 were reviewed. The six cases of possible gravity-drain paths using the water of the refueling water storage tank (RWST) were identified and the thermal hydraulic analyses were performed using RELAP5/MOD3.2 code. The core cooling capability was dependent on the gravity-drain paths and the drain rate. In the cases with the injection path and opening on the different leg side, the system was well depressurized after gravity-injection and the core boiling was successfully prevented for a long-term transient. However, in the cases with the injection path and opening on the cold leg side, the core coolant continued boiling although the system pressure remains atmospheric after gravity-injection because the cold water injected from the RWST was bypassed the core region. In the cases with the higher pressurizer opening than the RWST water level, the system was also pressurized by the water-hold in the pressurizer and the core was uncovered because the gravity-injection from the RWST stopped due to the high system pressure. In addition, from the sensitivity study on the gravity-injection flow rates, it was found that about 54 kg/s of RWST drain rate was required to maintain the core cooling. Those analysis results would provide useful information to operators coping with the event

  20. Aseismic study of high temperature gas-cooled reactor core with block-type fuel, 3

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1985-01-01

    A two-dimensional horizontal seismic experiment with single axis and simultaneous two-axes excitations was performed to obtain the core seismic design data on the block-type high temperature gas-cooled reactor. Effects of excitation directions and core side support stiffness on characteristics of core displacements and reaction forces of support were revealed. The values of the side reaction forces are the largest in the excitation of flat-to-flat of hexagonal block. Preload from the core periphery to the core center are effective to decrease core displacements and side reaction forces. (author)

  1. THE GROWTH OF COOL CORES AND EVOLUTION OF COOLING PROPERTIES IN A SAMPLE OF 83 GALAXY CLUSTERS AT 0.3 < z < 1.2 SELECTED FROM THE SPT-SZ SURVEY

    Energy Technology Data Exchange (ETDEWEB)

    McDonald, M.; Bautz, M. W. [Kavli Institute for Astrophysics and Space Research, Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States); Benson, B. A.; Bleem, L. E.; Carlstrom, J. E.; Chang, C. L.; Crawford, T. M.; Crites, A. T. [Kavli Institute for Cosmological Physics, University of Chicago, 5640 South Ellis Avenue, Chicago, IL 60637 (United States); Vikhlinin, A.; Stalder, B.; Ashby, M. L. N.; Bayliss, M. [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); De Haan, T. [Department of Physics, McGill University, 3600 Rue University, Montreal, Quebec H3A 2T8 (Canada); Lin, H. W. [Caddo Parish Magnet High School, Shrevport, LA 71101 (United States); Aird, K. A. [University of Chicago, 5640 South Ellis Avenue, Chicago, IL 60637 (United States); Bocquet, S.; Desai, S. [Department of Physics, Ludwig-Maximilians-Universitaet, Scheinerstr. 1, D-81679 Muenchen (Germany); Brodwin, M. [Department of Physics and Astronomy, University of Missouri, 5110 Rockhill Road, Kansas City, MO 64110 (United States); Cho, H. M. [NIST Quantum Devices Group, 325 Broadway Mailcode 817.03, Boulder, CO 80305 (United States); Clocchiatti, A., E-mail: mcdonald@space.mit.edu [Departamento de Astronomia y Astrosifica, Pontificia Universidad Catolica (Chile); and others

    2013-09-01

    We present first results on the cooling properties derived from Chandra X-ray observations of 83 high-redshift (0.3 < z < 1.2) massive galaxy clusters selected by their Sunyaev-Zel'dovich signature in the South Pole Telescope data. We measure each cluster's central cooling time, central entropy, and mass deposition rate, and compare these properties to those for local cluster samples. We find no significant evolution from z {approx} 0 to z {approx} 1 in the distribution of these properties, suggesting that cooling in cluster cores is stable over long periods of time. We also find that the average cool core entropy profile in the inner {approx}100 kpc has not changed dramatically since z {approx} 1, implying that feedback must be providing nearly constant energy injection to maintain the observed ''entropy floor'' at {approx}10 keV cm{sup 2}. While the cooling properties appear roughly constant over long periods of time, we observe strong evolution in the gas density profile, with the normalized central density ({rho}{sub g,0}/{rho}{sub crit}) increasing by an order of magnitude from z {approx} 1 to z {approx} 0. When using metrics defined by the inner surface brightness profile of clusters, we find an apparent lack of classical, cuspy, cool-core clusters at z > 0.75, consistent with earlier reports for clusters at z > 0.5 using similar definitions. Our measurements indicate that cool cores have been steadily growing over the 8 Gyr spanned by our sample, consistent with a constant, {approx}150 M{sub Sun} yr{sup -1} cooling flow that is unable to cool below entropies of 10 keV cm{sup 2} and, instead, accumulates in the cluster center. We estimate that cool cores began to assemble in these massive systems at z{sub cool}=1.0{sup +1.0}{sub -0.2}, which represents the first constraints on the onset of cooling in galaxy cluster cores. At high redshift (z {approx}> 0.75), galaxy clusters may be classified as ''cooling flows

  2. SCORPIO - VVER core surveillance system

    International Nuclear Information System (INIS)

    Zalesky, K.; Svarny, J.; Novak, L.; Rosol, J.; Horanes, A.

    1997-01-01

    The Halden Project has developed the core surveillance system SCORPIO which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. So far the system has only been implemented on western PWRs but the basic concept is applicable to a wide range of reactor including WWERs. The main differences between WWERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a WWER version of SCORPIO was initiated in cooperation with the Nuclear Research Institute at Rez and industry partners in the Czech Republic. The first system will be installed at the Dukovany NPP. (author)

  3. A fast converging CFD model for thermal hydraulic analysis of gas cooled reactor cores

    International Nuclear Information System (INIS)

    Chen, Gary; Anghaie, Samim

    1999-01-01

    A computational fluid dynamics (CFD) approach to the solution of Navier-Stokes equations for the thermal and flow fields of gas cooled reactor cores is presented. An implicit-explicit MacCormack method based on finite volume discretization scheme, in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve axisymmetric, thin-layer Navier-Stokes equations. This numerical method requires only the inversion of block bidiagonal systems rather than block tridiagonal systems, thus yielding savings in computer time and storage requirements. A two-layer algebraic eddy viscosity turbulence model is used in this study. The effects of turbulence are simulated in terms of the eddy viscosity coefficient, which is calculated for an inner and an outer region separately. An enthalpy-rebalancing scheme is implemented to allow the convergence solutions to be obtained with the application of a wall heat flux. The detailed computational analysis developed in this work is used to evaluate many different Nusselt number equations, property corrections, and axial distance corrections. The calculation based on this CFD model is compared with other published results. The good agreement indicates the usefulness of the presented model for the prediction of flow and temperature distributions for gas cooled reactor cores. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. Supporting system for the core restraint of nuclear reactors

    International Nuclear Information System (INIS)

    Kaser, A.

    1973-01-01

    The core restraint of water cooled nuclear reactors which is needed to direct the flow of the coolant through the core can be manufactured only in a moderate wall thickness. Thus, the majority of the loads have to be transmitted to the core barrel which is more rigid. The patent refers to a system of circumferential and vertical support members most of which are free to move relatively to each other, thus reducing thermal stresses during operation. (P.K.)

  6. Cooling system upon reactor isolation

    International Nuclear Information System (INIS)

    Yamamoto, Kohei; Oda, Shingo; Miura, Satoshi

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    1978-09-01

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

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

    International Nuclear Information System (INIS)

    1977-09-01

    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 [de

  9. Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core

    International Nuclear Information System (INIS)

    Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

    2003-01-01

    This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction

  10. Passive decay heat removal by sump cooling after core meltdown

    International Nuclear Information System (INIS)

    Knebel, J.U.; Mueller, U.

    1996-01-01

    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 a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives first measurement results of the 1:20 linearly scaled plane two-dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototype conditions

  11. Elastocaloric cooling materials and systems

    Science.gov (United States)

    Takeuchi, Ichiro

    2015-03-01

    We are actively pursuing applications of thermoelastic (elastocaloric) cooling using shape memory alloys. Latent heat associated with martensitic transformation of shape memory alloys can be used to run cooling cycles with stress-inducing mechanical drives. The coefficient of performance of thermoelastic cooling materials can be as high as 11 with the directly measured DT of around 17 °C. Depending on the stress application mode, the number of cycles to fatigue can be as large as of the order of 105. Efforts to design and develop thermoelastic alloys with long fatigue life will be discussed. The current project at the University of Maryland is focused on development of building air-conditioners, and at Maryland Energy and Sensor Technologies, smaller scale commercial applications are being pursued. This work is carried out in collaboration with Jun Cui, Yiming Wu, Suxin Qian, Yunho Hwang, Jan Muehlbauer, and Reinhard Radermacher, and it is funded by the ARPA-E BEETIT program and the State of Maryland.

  12. Heat transfer calculations on the KNK II emergency cooling system

    International Nuclear Information System (INIS)

    Vossebrecker, H.; Groenefeld, G.

    1976-12-01

    The Licensing Authority had demanded that in case of the change of the KNK thermal core into a fast core the decay heat removal system must be improved by a diverse and spatially separated emergency cooling system. In order to meet this requirement an existing nitrogen system of the facility is extended in such a manner that the decay heat will be removed by a nitrogen flow passing through the gap between reactor vessel and guard vessel. The heat transport from the core to the vessel is accomplished by natural convection flow rates which are generated by density differences between the hot core subassemblies, the reflector subassemblies and other passages between the upper and the lower plenum. The calculations show that the maximum temperatures in the core do not reach the sodium boiling-point. The maximum vessel temperature is 673 deg. C. In this report the function of the emergency cooling system and the methods of calculation are described, the input data and the results are stated and it is shown that the calculated temperatures are conservative [de

  13. Overheating preventive system for reactor core fuels

    International Nuclear Information System (INIS)

    Ito, Daiju

    1981-01-01

    Purpose: To ensure the cooling function of reactor water in a cooling system in case of erroneous indication or misoperation by reliable temperature measurement for fuels and actuating relays through the conversion output obtained therefrom. Constitution: Thermometers are disposed laterally and vertically in a reactor core in contact with core fuels so as to correspond to the change of status in the reactor core. When there is a high temperature signal issued from one of the thermometers or one of conversion circuits, the function of relay contacts does not provide the closed state as a whole. When high temperature signals are issued from two or more thermometers of conversion circuits from independent OR circuits, the function of the relay contacts provides the closure state as a whole. Consequently, in the use of 2-out of 3-circuits, the entire closure state, that is, the misoperation of the relay contacts for the thermometer or the conversion circuits can be avoided. In this way, by the application of the output from the conversion circuits to the logic circuit and, in turn, application of the output therefrom to the relay groups in 2-out of 3-constitution, the reactor safety can be improved. (Horiuchi, T.)

  14. Cooling Grapple System for FMEF hot cell

    International Nuclear Information System (INIS)

    Semmens, L.S.; Frandsen, G.B.; Tome, R.

    1983-01-01

    A Cooling Grapple System was designed and built to handle fuel assemblies within the FMEF hot cell. The variety of functions for which it is designed makes it unique from grapples presently in use. The Cooling Grapple can positively grip and transport assemblies vertically, retrieve assemblies from molten sodium where six inches of grapple tip is submerged, cool 7 kw assemblies in argon, and service an in-cell area of 372 m 2 (4000 ft 2 ). Novel and improved operating and maintenance features were incorporated in the design including a shear pin and mechanical catcher system to prevent overloading the grapple while allowing additional reaction time for crane shutdown

  15. The ATLAS IBL CO2 Cooling System

    CERN Document Server

    Verlaat, Bartholomeus; The ATLAS collaboration

    2016-01-01

    The Atlas Pixel detector has been equipped with an extra B-layer in the space obtained by a reduced beam pipe. This new pixel detector called the ATLAS Insertable B-Layer (IBL) is installed in 2014 and is operational in the current ATLAS data taking. The IBL detector is cooled with evaporative CO2 and is the first of its kind in ATLAS. The ATLAS IBL CO2 cooling system is designed for lower temperature operation (<-35⁰C) than the previous developed CO2 cooling systems in High Energy Physics experiments. The cold temperatures are required to protect the pixel sensors for the high expected radiation dose up to 550 fb^-1 integrated luminosity. This paper describes the design, development, construction and commissioning of the IBL CO2 cooling system. It describes the challenges overcome and the important lessons learned for the development of future systems which are now under design for the Phase-II upgrade detectors.

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

    International Nuclear Information System (INIS)

    Ball, S.J.

    1991-10-01

    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

  17. Dry cooling systems with plastic surfaces

    International Nuclear Information System (INIS)

    Roma, Carlo; Leonelli, Vincenzo

    1975-01-01

    Research and experiments made on dry cooling systems with plastic surfaces are described. The demonstration program planned in Italy for a 100Gcal/h dry cooling system is exposed, and an installation intended for a large 1300Mwe nuclear power station is described with reference to the assembly (exploitation and maintenance included). The performance and economic data relating to this installation are also exposed [fr

  18. SMART core protection system design

    International Nuclear Information System (INIS)

    Lee, J. K.; Park, H. Y.; Koo, I. S.; Park, H. S.; Kim, J. S.; Son, C. H.

    2003-01-01

    SMART COre Protection System(SCOPS) is designed with real-tims Digital Signal Processor(DSP) board and Network Interface Card(NIC) board. SCOPS has a Control Rod POSition (CRPOS) software module while Core Protection Calculator System(CPCS) consists of Core Protection Calculators(CPCs) and Control Element Assembly(CEA) Calculators(CEACs) in the commercial nuclear plant. It's not necessary to have a independent cabinets for SCOPS because SCOPS is physically very small. Then SCOPS is designed to share the cabinets with Plant Protection System(PPS) of SMART. Therefor it's very easy to maintain the system because CRPOS module is used instead of the computer with operating system

  19. 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.

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

    International Nuclear Information System (INIS)

    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

  1. Understanding aging in containment cooling systems

    International Nuclear Information System (INIS)

    Lofaro, R.J.

    1993-01-01

    A study has been performed to assess the effects of aging in nuclear power plant containment cooling systems. Failure records from national databases, as well as plant specific data were reviewed and analyzed to identify aging characteristics for this system. The predominant aging mechanisms were determined, along with the most frequently failed components and their associated failure modes. This paper discusses the aging mechanisms present in the containment spray system and the containment fan cooler system, which are two systems used to provide the containment cooling function. The failure modes, along with the relative frequency of each is also discussed

  2. New Protective Measures for Cooling Systems

    International Nuclear Information System (INIS)

    Carter, D. Anthony; Nonohue, Jonh M.

    1974-01-01

    Cooling water treatments have been updated and improved during the last few years. Particularly important are the nontoxic programs which conform plant cooling water effluents to local water quality standards without expenditures for capital equipment. The relationship between scaling and corrosion in natural waters has been recognized for many years. This relationship is the basis for the Langelier Saturation Index control method which was once widely applied to reduce corrosion in cooling water systems. It used solubility characteristics to maintain a very thin deposit on metal surfaces for preventing corrosion. This technique was rarely successful. That is, the solubility of calcium carbonate and most other inorganic salts depends on temperature. If good control exists on cold surfaces, excessive deposition results on the heat transfer tubes. Also, because water characteristic normally vary in a typical cooling system, precise control of scaling at both hot and cold surfaces is virtually impossible

  3. Cooling system with automated seasonal freeze protection

    Science.gov (United States)

    Campbell, Levi A.; Chu, Richard C.; David, Milnes P.; Ellsworth, Jr., Michael J.; Iyengar, Madhusudan K.; Simons, Robert E.; Singh, Prabjit; Zhang, Jing

    2016-05-24

    An automated multi-fluid cooling system and method are provided for cooling an electronic component(s). The cooling system includes a coolant loop, a coolant tank, multiple valves, and a controller. The coolant loop is at least partially exposed to outdoor ambient air temperature(s) during normal operation, and the coolant tank includes first and second reservoirs containing first and second fluids, respectively. The first fluid freezes at a lower temperature than the second, the second fluid has superior cooling properties compared with the first, and the two fluids are soluble. The multiple valves are controllable to selectively couple the first or second fluid into the coolant in the coolant loop, wherein the coolant includes at least the second fluid. The controller automatically controls the valves to vary first fluid concentration level in the coolant loop based on historical, current, or anticipated outdoor air ambient temperature(s) for a time of year.

  4. 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.

  5. Investigations on sump cooling after core melt down

    International Nuclear Information System (INIS)

    Knebel, J.U.

    1995-01-01

    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

  6. Rust Inhibitor And Fungicide For Cooling Systems

    Science.gov (United States)

    Adams, James F.; Greer, D. Clay

    1988-01-01

    Mixture of benzotriazole, benzoic acid, and fungicide prevents growth of rust and fungus. Water-based cooling mixture made from readily available materials prevents formation of metallic oxides and growth of fungi in metallic pipes. Coolant remains clear and does not develop thick sludge tending to collect in low points in cooling systems with many commercial rust inhibitors. Coolant compatible with iron, copper, aluminum, and stainless steel. Cannot be used with cadmium or cadmium-plated pipes.

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

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    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)

  8. Controlled cooling of an electronic system for reduced energy consumption

    Science.gov (United States)

    David, Milnes P.; Iyengar, Madhusudan K.; Schmidt, Roger R.

    2016-08-09

    Energy efficient control of a cooling system cooling an electronic system is provided. The control includes automatically determining at least one adjusted control setting for at least one adjustable cooling component of a cooling system cooling the electronic system. The automatically determining is based, at least in part, on power being consumed by the cooling system and temperature of a heat sink to which heat extracted by the cooling system is rejected. The automatically determining operates to reduce power consumption of the cooling system and/or the electronic system while ensuring that at least one targeted temperature associated with the cooling system or the electronic system is within a desired range. The automatically determining may be based, at least in part, on one or more experimentally obtained models relating the targeted temperature and power consumption of the one or more adjustable cooling components of the cooling system.

  9. Controlled cooling of an electronic system for reduced energy consumption

    Energy Technology Data Exchange (ETDEWEB)

    David, Milnes P.; Iyengar, Madhusudan K.; Schmidt, Roger R.

    2018-01-30

    Energy efficient control of a cooling system cooling an electronic system is provided. The control includes automatically determining at least one adjusted control setting for at least one adjustable cooling component of a cooling system cooling the electronic system. The automatically determining is based, at least in part, on power being consumed by the cooling system and temperature of a heat sink to which heat extracted by the cooling system is rejected. The automatically determining operates to reduce power consumption of the cooling system and/or the electronic system while ensuring that at least one targeted temperature associated with the cooling system or the electronic system is within a desired range. The automatically determining may be based, at least in part, on one or more experimentally obtained models relating the targeted temperature and power consumption of the one or more adjustable cooling components of the cooling system.

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

    International Nuclear Information System (INIS)

    Kumpf, H.

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Downey, G.L.

    1988-01-01

    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)

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

    International Nuclear Information System (INIS)

    Chambadal, P.; Pascal, M.

    1955-01-01

    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.) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

    Downey, G L

    1988-05-01

    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.

  14. Preoperational test report, recirculation condenser cooling systems

    International Nuclear Information System (INIS)

    Clifton, F.T.

    1997-01-01

    This represents a preoperational test report for Recirculation Condenser Systems, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The four system provide condenser cooling water for vapor space cooling of tanks AY1O1, AY102, AZ1O1, AZ102. Each system consists of a valved piping loop, a pair of redundant recirculation pumps, a closed-loop evaporative cooling tower, and supporting instrumentation; equipment is located outside the farm on concrete slabs. Piping is routed to the each ventilation condenser inside the farm via below-grade concrete trenches. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System

  15. Preoperational test report, recirculation condenser cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Clifton, F.T.

    1997-11-04

    This represents a preoperational test report for Recirculation Condenser Systems, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The four system provide condenser cooling water for vapor space cooling of tanks AY1O1, AY102, AZ1O1, AZ102. Each system consists of a valved piping loop, a pair of redundant recirculation pumps, a closed-loop evaporative cooling tower, and supporting instrumentation; equipment is located outside the farm on concrete slabs. Piping is routed to the each ventilation condenser inside the farm via below-grade concrete trenches. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

  16. SCORPIO - WWER core surveillance system

    International Nuclear Information System (INIS)

    Hornaes, Arne; Bodal, Terje; Sunde, Svein; Zalesky, K.; Lehman, M.; Pecka, M.; Svarny, J.; Krysl, V.; Juzova, Z.; Sedlak, A.; Semmler, M.

    1998-01-01

    The Institut for energiteknikk has developed the core surveillance system SCORPIO, which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety, as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. The system has been implemented on western PWRs, but the basic concept is applicable to a wide range of reactors including WWERs. The main differences between WWERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a WWER version of SCORPIO has been done in co-operation with the Nuclear Research Institute Rez, and industry partners in the Czech Republic. The first system is installed at Dukovany NPP, where the Site Acceptance Test was completed 6. March 1998.(Authors)

  17. SCORPIO - VVER core surveillance system

    International Nuclear Information System (INIS)

    Hornaes, A.; Bodal, T.; Sunde, S.

    1998-01-01

    The Institutt for energiteknikk has developed the core surveillance system SCORPIO, which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators, which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety, as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. The system has been implemented on western PWRs, but the basic concept is applicable to a wide range of reactors including VVERs. The main differences between VVERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a VVER version of SCORPIO has been done in co-operation with the Nuclear Research Institute Rez, and industry partners in the Czech Republic. The first system is installed at Dukovany NPP, where the Site Acceptance Test was completed 6. March 1998.(author)

  18. Provision of reliable core cooling in vessel-type boiling reactors

    International Nuclear Information System (INIS)

    Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

    1987-01-01

    Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

  19. Gas-cooled reactor power systems for space

    International Nuclear Information System (INIS)

    Walter, C.E.

    1987-01-01

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

  20. System for Cooling of Electronic Components

    Science.gov (United States)

    Vasil'ev, L. L.; Grakovich, L. P.; Dragun, L. A.; Zhuravlev, A. S.; Olekhnovich, V. A.; Rabetskii, M. I.

    2017-01-01

    Results of computational and experimental investigations of heat pipes having a predetermined thermal resistance and a system based on these pipes for air cooling of electronic components and diode assemblies of lasers are presented. An efficient compact cooling system comprising heat pipes with an evaporator having a capillary coating of a caked copper powder and a condenser having a developed outer finning, has been deviced. This system makes it possible to remove, to the ambient air, a heat flow of power more than 300 W at a temperature of 40-50°C.

  1. MULTIFUNCTIONAL SOLAR SYSTEMS FOR HEATING AND COOLING

    Directory of Open Access Journals (Sweden)

    Doroshenko A.V.

    2010-12-01

    Full Text Available The basic circuits of multifunctional solar systems of air drainage, heating (hot water supply and heating, cooling and air conditioning are developed on the basis of open absorption cycle with a direct absorbent regeneration. Basic decisions for new generation of gas-liquid solar collectors are developed. Heat-mass-transfer apparatus included in evaporative cooling system, are based on film interaction of flows of gas and liquid and in them, for the creation of nozzle, multi-channel structures from polymeric materials and porous ceramics are used. Preliminary analysis of multifunctional systems possibilities is implemented.

  2. Status of helium-cooled nuclear power systems. [Development potential

    Energy Technology Data Exchange (ETDEWEB)

    Melese-d' Hospital, G.; Simnad, M

    1977-09-01

    Helium-cooled nuclear power systems offer a great potential for electricity generation when their long-term economic, environmental, conservation and energy self-sufficiency features are examined. The high-temperature gas-cooled reactor (HTGR) has the unique capability of providing high-temperature steam for electric power and process heat uses and/or high-temperature heat for endothermic chemical reactions. A variation of the standard steam cycle HTGR is one in which the helium coolant flows directly from the core to one or more closed cycle gas turbines. The effective use of nuclear fuel resources for electric power and nuclear process heat will be greatly enhanced by the gas-cooled fast breeder reactor (GCFR) currently being developed. A GCFR using thorium in the radial blanket could generate sufficient U-233 to supply the fuel for three HTGRs, or enough plutonium from a depleted uranium blanket to fuel a breeder economy expanding at about 10% per year. The feasibility of utilizing helium to cool a fusion reactor is also discussed. The status of helium-cooled nuclear energy systems is summarized as a basis for assessing their prospects. 50 references.

  3. CHANDRA OBSERVATION OF ABELL 1142: A COOL-CORE CLUSTER LACKING A CENTRAL BRIGHTEST CLUSTER GALAXY?

    Energy Technology Data Exchange (ETDEWEB)

    Su, Yuanyuan; Weeren, Reinout van [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Buote, David A. [Department of Physics and Astronomy, University of California, Irvine, 4129 Frederick Reines Hall, Irvine, CA 92697 (United States); Gastaldello, Fabio, E-mail: yuanyuan.su@cfa.harvard.edu [INAF-IASF-Milano, Via E. Bassini 15, I-20133 Milano (Italy)

    2016-04-10

    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{sup −1}. This ongoing merger may have shock-heated the ICM from ≈2 keV to above 3 keV, which would explain the anomalous L{sub X}–T{sub X} 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.

  4. Slab cooling system design using computer simulation

    NARCIS (Netherlands)

    Lain, M.; Zmrhal, V.; Drkal, F.; Hensen, J.L.M.

    2007-01-01

    For a new technical library building in Prague computer simulations were carried out to help design of slab cooling system and optimize capacity of chillers. In the paper is presented concept of new technical library HVAC system, the model of the building, results of the energy simulations for

  5. Radiant Heating and Cooling Systems. Part two

    DEFF Research Database (Denmark)

    Kim, Kwan Woo; Olesen, Bjarne W.

    2015-01-01

    Control of the heating and cooling system needs to be able to maintain the indoor temperatures within the comfort range under the varying internal loads and external climates. To maintain a stable thermal environment, the control system needs to maintain the balance between the heat gain...

  6. Shivering heat production and body fat protect the core from cooling during body immersion, but not during head submersion: a structural equation model.

    Science.gov (United States)

    Pretorius, Thea; Lix, Lisa; Giesbrecht, Gordon

    2011-03-01

    Previous studies showed that core cooling rates are similar when only the head or only the body is cooled. Structural equation modeling was used on data from two cold water studies involving body-only, or whole body (including head) cooling. Exposure of both the body and head increased core cooling, while only body cooling elicited shivering. Body fat attenuates shivering and core cooling. It is postulated that this protection occurs mainly during body cooling where fat acts as insulation against cold. This explains why head cooling increases surface heat loss with only 11% while increasing core cooling by 39%. Copyright © 2011 Elsevier Ltd. All rights reserved.

  7. Smart Cooling Controlled System Exploiting Photovoltaic Renewable Energy Systems

    Directory of Open Access Journals (Sweden)

    Ahmad Atieh

    2018-03-01

    Full Text Available A smart cooling system to control the ambient temperature of a premise in Amman, Jordan, is investigated and implemented. The premise holds 650 people and has 14 air conditioners with the cooling capacity ranging from 3 to 5 ton refrigerant (TR each. The control of the cooling system includes implementing different electronics circuits that are used to sense the ambient temperature and humidity, count the number of people in the premise and then turn ON/OFF certain air conditioner(s. The data collected by different electronic circuits are fed wirelessly to a microcontroller, which decides which air conditioner will be turned ON/OFF, its location and its desired set cooling temperature. The cooling system is integrated with an on-grid solar photovoltaic energy system to minimize the operational cost of the overall cooling system.

  8. Passive cooling systems in power reactors

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  9. Power plant cooling systems: trends and challenges

    International Nuclear Information System (INIS)

    Rittenhouse, R.C.

    1979-01-01

    A novel design for an intake and discharge system at the Belle River plant is described followed by a general discussion of water intake screens and porous dikes for screening fish and zooplankton. The intake system for the San Onofre PWR plant is described and the state regulations controlling the use of water for power plants is discussed. The use of sewage effluent as a source of cooling water is mentioned with reference to the Palo Verde plant. Progress in dry cooling and a new wet/dry tower due to be installed at the San Juan plant towards the end of this year, complete the survey

  10. Stochastic cooling system in COSY

    International Nuclear Information System (INIS)

    Brittner, P.; Hacker, H.U.; Prasuhn, D.; Schug, G.; Singer, H.; Spiess, W.; Stassen, R.

    1994-01-01

    The stochastic cooler system in COSY is designed for proton kinetic energies between 0.8 and 2.5 GeV. Fabrication of the mechanical parts of the system is going on. Test results of the prototype measurements as well as data of the active RF-compontens are presented. (orig.)

  11. Stochastic cooling system in COSY

    Energy Technology Data Exchange (ETDEWEB)

    Brittner, P [Forschungszentrum Juelich GmbH (Germany); Hacker, H U [Forschungszentrum Juelich GmbH (Germany); Prasuhn, D [Forschungszentrum Juelich GmbH (Germany); Schug, G [Forschungszentrum Juelich GmbH (Germany); Singer, H [Forschungszentrum Juelich GmbH (Germany); Spiess, W [Forschungszentrum Juelich GmbH (Germany); Stassen, R [Forschungszentrum Juelich GmbH (Germany)

    1994-09-01

    The stochastic cooler system in COSY is designed for proton kinetic energies between 0.8 and 2.5 GeV. Fabrication of the mechanical parts of the system is going on. Test results of the prototype measurements as well as data of the active RF-compontens are presented. (orig.)

  12. Core design characteristics of the hyper system

    International Nuclear Information System (INIS)

    Yonghee, Kim; Won-Seok, Park; Hill, R.N.

    2003-01-01

    In Korea, an accelerator-driven system (ADS) called HYPER (Hybrid Power Extraction Reactor) is being studied for the transmutation of the radioactive wastes. HYPER is a 1000 MWth lead-bismuth eutectic (LBE)-cooled ADS. In this paper, the neutronic design characteristics of HYPER are described and its transmutation performances are assessed for an equilibrium cycle. The core is loaded with a ductless fuel assembly containing transuranics (TRU) dispersion fuel pins. In HYPER, a relatively high core height, 160 cm, is adopted to maximize the multiplication efficiency of the external source. In the ductless fuel assembly, 13 non-fuel rods are used as tie rods to maintain the mechanical integrity of assembly. As the reflector material, pure lead is used to improve the neutron economy and to minimise the generation of radioactive materials. In HYPER, to minimise the burn-up reactivity swing, a B 4 C burnable absorber is employed. For efficient depletion of the B-10 absorber, the burnable absorber is loaded only in the axially-central part (92 cm long) of the 13 tie rods of each assembly. In the current design, the amount of the B 4 C absorber was determined such that the burn-up reactivity swing is about 3.0% Δk. The long-lived fission products (LLFPs) 99 Tc and 129 I are also transmuted in the HYPER core such that their supporting ratios are equal to that of the TRUs. A heterogeneous LLFP transmutation in the reflector zone has been analysed in this work. A unique feature of the HYPER system is that it has an auxiliary core shutdown system, independent of the accelerator shutdown system. It has been shown that a cylindrical B 4 C absorber between the target and fuel blanket can drastically reduce the fission power even without shutting off the accelerator power. (author)

  13. Atmospheric impacts of evaporative cooling systems

    International Nuclear Information System (INIS)

    Carson, J.E.

    1976-10-01

    The report summarizes available information on the effects of various power plant cooling systems on the atmosphere. While evaporative cooling systems sharply reduce the biological impacts of thermal discharges in water bodies, they create (at least, for heat-release rates comparable to those of two-unit nuclear generating stations) atmospheric changes. For an isolated site such as required for a nuclear power plant, these changes are rather small and local, and usually environmentally acceptable. However, one cannot say with certainty that these effects will remain small as the number of reactors on a given site increases. There must exist a critical heat load for a specific site which, if exceeded, can create its own weather patterns, and thus create inadvertent weather changes such as rain and snow, severe thunderstorms, and tornadoes. Because proven mathematical models are not available, it is not now possible to forecast precisely the extent and frequency of the atmospheric effects of a particular heat-dissipation system at a particular site. Field research on many aspects of cooling system operation is needed in order to document and quantify the actual atmospheric changes caused by a given cooling system and to provide the data needed to develop and verify mathematical and physical models. The more important topics requiring field study are plume rise, fogging and icing (from certain systems), drift emission and deposition rates, chemical interactions, cloud and precipitation formation and critical heat-release rates

  14. The ATLAS IBL CO2 Cooling System

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00237783; The ATLAS collaboration; Zwalinski, L.; Bortolin, C.; Vogt, S.; Godlewski, J.; Crespo-Lopez, O.; Van Overbeek, M.; Blaszcyk, T.

    2017-01-01

    The ATLAS Pixel detector has been equipped with an extra B-layer in the space obtained by a reduced beam pipe. This new pixel detector called the ATLAS Insertable B-Layer (IBL) is installed in 2014 and is operational in the current ATLAS data taking. The IBL detector is cooled with evaporative CO2 and is the first of its kind in ATLAS. The ATLAS IBL CO2 cooling system is designed for lower temperature operation (<-35⁰C) than the previous developed CO2 cooling systems in High Energy Physics experiments. The cold temperatures are required to protect the pixel sensors for the high expected radiation dose up to 550 fb^-1 integrated luminosity.

  15. Advanced Small-Safe Long-Life Lead Cooled Reactor Cores for Future Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Hyeong; Hong, Ser Gi [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    One of the reasons for use of the lead or lead-bismuth alloy coolants is the high boiling temperature that avoids the possibility of coolant voiding. Also, these coolants are compatible with air, steam, and water. Therefore, intermediate coolant loop is not required as in the sodium cooled reactors 3. Lead is considered to be more attractive coolant than lead-bismuth alloy because of its higher availability, lower price, and much lower amount of polonium activity by factor of 104 relatively to lead. On the other hand, lead has higher melting temperature of 601K than that of lead-bismuth (398K), which narrows the operating temperature range and also leads to the possibility of freezing and blockage in fresh cores. Neutronically, the lead and lead-bismuth have very similar characteristics to each other. The lead-alloy coolants have lower moderating power and higher scattering without increasing moderation for neutrons below 0.5MeV, which reduces the leakage of the neutrons through the core and provides an excellent reflecting capability for neutrons. Due to the above features of lead or lead-alloy coolants, there have been lots of studies on the small lead cooled core designs. In this paper, small-safe long-life lead cooled reactor cores having high discharge burnup are designed and neutronically analyzed.. The cores considered in this work rates 110MWt (36.7MWe). In this work, the long-life with high discharge burnup was achieved by using thorium or depleted uranium blanket loaded in the central region of the core. Also, we considered a reference core having no blanket for the comparison. This paper provides the detailed neutronic analyses for these small long-life cores and the detailed analyses of the reactivity coefficients and the composition changes in blankets. The results of the core design and analyses show that our small long-life cores can be operated without refueling over their long-lives longer than 45EFPYs (Effective Full Power Year). In this work

  16. A design method to isothermalize the core of high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takano, M.; Sawa, K.

    1987-01-01

    A practical design method is developed to isothermalize the core of block-type high-temperature gas-cooled reactors (HTGRs). Isothermalization plays an important role in increasing the design margin on fuel temperature. In this method, the fuel enrichment and the size and boron content of the burnable poison rod are determined over the core blockwise so that the axially exponential and radially flat power distribution are kept from the beginning to the end of core life. The method enables conventional HTGRs to raise the outlet gas temperature without increasing the maximum fuel temperature

  17. Real time thermal hydraulic model for high temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Sui Zhe; Sun Jun; Ma Yuanle; Zhang Ruipeng

    2013-01-01

    A real-time thermal hydraulic model of the reactor core was described and integrated into the simulation system for the high temperature gas-cooled pebble bed reactor nuclear power plant, which was developed in the vPower platform, a new simulation environment for nuclear and fossil power plants. In the thermal hydraulic model, the helium flow paths were established by the flow network tools in order to obtain the flow rates and pressure distributions. Meanwhile, the heat structures, representing all the solid heat transfer elements in the pebble bed, graphite reflectors and carbon bricks, were connected by the heat transfer network in order to solve the temperature distributions in the reactor core. The flow network and heat transfer network were coupled and calculated in real time. Two steady states (100% and 50% full power) and two transients (inlet temperature step and flow step) were tested that the quantitative comparisons of the steady results with design data and qualitative analysis of the transients showed the good applicability of the present thermal hydraulic model. (authors)

  18. Operational cost minimization in cooling water systems

    Directory of Open Access Journals (Sweden)

    Castro M.M.

    2000-01-01

    Full Text Available In this work, an optimization model that considers thermal and hydraulic interactions is developed for a cooling water system. It is a closed loop consisting of a cooling tower unit, circulation pump, blower and heat exchanger-pipe network. Aside from process disturbances, climatic fluctuations are considered. Model constraints include relations concerning tower performance, air flowrate requirement, make-up flowrate, circulating pump performance, heat load in each cooler, pressure drop constraints and climatic conditions. The objective function is operating cost minimization. Optimization variables are air flowrate, forced water withdrawal upstream the tower, and valve adjustment in each branch. It is found that the most significant operating cost is related to electricity. However, for cooled water temperatures lower than a specific target, there must be a forced withdrawal of circulating water and further makeup to enhance the cooling tower capacity. Additionally, the system is optimized along the months. The results corroborate the fact that the most important variable on cooling tower performance is not the air temperature itself, but its humidity.

  19. Void worths in subcritical cores cooled by lead-bismuth

    International Nuclear Information System (INIS)

    Wallenius, Janne; Tucek, Kamil; Gudowski, Waclaw

    2001-01-01

    The introduction lead-bismuth coolant in accelerator driven transmutation systems (ADS) was: good neutron economy (higher source efficiency); natural circulation possible (decay heat removal); synergy with spallation target (simplified coolant management); high temperature of boiling (larger overpower margin); smaller void worths (operation at higher k-values). This paper deals with different aspects of the void worths in JAERI ADS

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

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    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 (YH 1.8 ) 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)

  1. Decontamination of primary cooling system

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake.

    1985-01-01

    Purpose: To effectively eliminate radioactivity accumulated in pipeways, equipments, etc in primary coolant circuits of BWR type power plants by utilizing ion displacement reactions. Method: The reactor pressure vessel is connected with a feedwater pipeway, steam pipeway and a recycling pipeway. The recycling pipeway is disposed with a recycling pump. A recycling by-pass line is branched from the recycling pipeway and disposed with a recycling system heat exchanger and chemical injection point. Water is filled in the primary coolant and heated 280 0 C. Then, while maintaining water at that temperature, non-radioactive cobalt ions are injected and circulated within the system, by which radioactivity accumulated in pipeways, equipments or the likes can effectively be removed. (Horiuchi, T.)

  2. Method of fabricating a cooled electronic system

    Science.gov (United States)

    Chainer, Timothy J; Gaynes, Michael A; Graybill, David P; Iyengar, Madhusudan K; Kamath, Vinod; Kochuparambil, Bejoy J; Schmidt, Roger R; Schultz, Mark D; Simco, Daniel P; Steinke, Mark E

    2014-02-11

    A method of fabricating a liquid-cooled electronic system is provided which includes an electronic assembly having an electronics card and a socket with a latch at one end. The latch facilitates securing of the card within the socket. The method includes providing a liquid-cooled cold rail at the one end of the socket, and a thermal spreader to couple the electronics card to the cold rail. The thermal spreader includes first and second thermal transfer plates coupled to first and second surfaces on opposite sides of the card, and thermally conductive extensions extending from end edges of the plates, which couple the respective transfer plates to the liquid-cooled cold rail. The extensions are disposed to the sides of the latch, and the card is securable within or removable from the socket using the latch without removing the cold rail or the thermal spreader.

  3. 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

  4. Turbine airfoil with laterally extending snubber having internal cooling system

    Science.gov (United States)

    Scribner, Carmen Andrew; Messmann, Stephen John; Marsh, Jan H.

    2016-09-06

    A turbine airfoil usable in a turbine engine and having at least one snubber with a snubber cooling system positioned therein and in communication with an airfoil cooling system is disclosed. The snubber may extend from the outer housing of the airfoil toward an adjacent turbine airfoil positioned within a row of airfoils. The snubber cooling system may include an inner cooling channel separated from an outer cooling channel by an inner wall. The inner wall may include a plurality of impingement cooling orifices that direct impingement fluid against an outer wall defining the outer cooling channel. In one embodiment, the cooling fluids may be exhausted from the snubber, and in another embodiment, the cooling fluids may be returned to the airfoil cooling system. Flow guides may be positioned in the outer cooling channel, which may reduce cross-flow by the impingement orifices, thereby increasing effectiveness.

  5. Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

    Directory of Open Access Journals (Sweden)

    Li Yuquan

    2017-02-01

    Full Text Available The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA, the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the

  6. Comparative experiments to assess the effects of accumulator nitrogen injection on passive core cooling during small break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Li, YuQuan; Hao, Botao; Zhong, Jia; Wan Nam [State Nuclear Power Technology R and D Center, South Park, Beijing Future Science and Technology City, Beijing (China)

    2017-02-15

    The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME)—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative

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

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    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 advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    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

  9. On synthesis and optimization of cooling water systems with multiple cooling towers

    CSIR Research Space (South Africa)

    Gololo, KV

    2011-01-01

    Full Text Available -1 On Synthesis and Optimization of Cooling Water Systems with Multiple Cooling Towers Khunedi Vincent Gololo?? and Thokozani Majozi*? ? Department of Chemical Engineering, University of Pretoria, Lynnwood Road, Pretoria, 0002, South Africa ? Modelling...

  10. Emergency cooling system for a liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Murata, Ryoichi; Fujiwara, Toshikatsu.

    1980-01-01

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

  11. Stochastic cooling with a double rf system

    International Nuclear Information System (INIS)

    Wei, Jie.

    1992-01-01

    Stochastic cooling for a bunched beam of hadrons stored in an accelerator with a double rf system of two different frequencies has been investigated. The double rf system broadens the spread in synchrotron-oscillation frequency of the particles when they mostly oscillate near the center of the rf bucket. Compared with the ease of a single rf system, the reduction rates of the bunch dimensions are significantly increased. When the rf voltage is raised, the reduction rate, instead of decreasing linearly, now is independent of the ratio of the bunch area to the bucket area. On the other hand, the spread in synchrotron-oscillation frequency becomes small with the double rf system, if the longitudinal oscillation amplitudes of the particles are comparable to the dimension of the rf bucket. Consequently, stochastic cooling is less effective when the bunch area is close to the bucket area

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

    International Nuclear Information System (INIS)

    Basset, F A; Cahill, F; Handrigan, G; DuCharme, M B; Cheung, S S

    2011-01-01

    Rectal (T re ), ear canal (T ear ) and esophageal (T es ) 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 m 2 ) 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 . m 2 . 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, T re was significantly lower (Δ = −1.0 ± 0.4 °C) from baseline values than from T ear (Δ = −0.3 ± 0.3 °C) and T es (Δ = −0.1 ± 0.3 °C). T re 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 T re 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

  13. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    International Nuclear Information System (INIS)

    Hunsbedt, A.; Boardman, C.E.

    1993-01-01

    A dual passive cooling system for liquid metal cooled nuclear fission reactors is described, comprising the combination of: a reactor vessel for containing a pool of liquid metal coolant with a core of heat generating fissionable fuel substantially submerged therein, a side wall of the reactor vessel forming an innermost first partition; a containment vessel substantially surrounding the reactor vessel in spaced apart relation having a side wall forming a second partition; a first baffle cylinder substantially encircling the containment vessel in spaced apart relation having an encircling wall forming a third partition; a guard vessel substantially surrounding the containment vessel and first baffle cylinder in spaced apart relation having a side wall forming a forth partition; a sliding seal at the top of the guard vessel edge to isolate the dual cooling system air streams; a second baffle cylinder substantially encircling the guard vessel in spaced part relationship having an encircling wan forming a fifth partition; a concrete silo substantially surrounding the guard vessel and the second baffle cylinder in spaced apart relation providing a sixth partition; a first fluid coolant circulating flow course open to the ambient atmosphere for circulating air coolant comprising at lent one down comer duct having an opening to the atmosphere in an upper area thereof and making fluid communication with the space between the guard vessel and the first baffle cylinder and at least one riser duct having an opening to the atmosphere in the upper area thereof and making fluid communication with the space between the first baffle cylinder and the containment vessel whereby cooling fluid air can flow from the atmosphere down through the down comer duct and space between the forth and third partitions and up through the space between the third and second partition and the riser duct then out into the atmosphere; and a second fluid coolant circulating flow

  14. Optimizing cooling systems in Egyptian arid urbans

    International Nuclear Information System (INIS)

    Medhat, Ahmed A.; Khalil, Essam E.

    2006-01-01

    Present study is devoted to climatic and site oriented investigations that were carried out in a new rural development in the Upper-Egypt. Bioclimatic classifications considered Upper Egypt region, near Sudan border, as a Hot and Dry climatic region. [1]. that is affected by solar heat intensities that can reach 900 W/m2 for a period ranged from 5-to-7 hours per day with the presence of study storms. Cooling season extends up to eight months per year having Upper-day-bulb temperature ranged from 400 degree centigrade - to - 470 degree centigrade while Lower-dry-bulb-temperature ranged from 280 degree centigrade - to - 320 degree centigrade with the relative humidity ranged from 10%-to-37% RH. [2]. Site surveys and field experimental and analyses of the commonly used cooling systems were investigated, evaluated and optimized for optimum indoor comfort conditions at efficient energy efficiency. [3]. Extensive analyses were performed based on Psychrometric formulae to evaluate the impact of energy consumptions related to different cooling systems such as direct expansion, chilled water, and evaporative systems. the present study enables the critical investigations of the influence of arid outdoor conditions and the required indoor thermal parameters on the energy efficiencies of HVAC-system. This work; focuses on the suggestion of suitable system that should be implemented by local energy codes in these arid urban.(Author)

  15. 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.

  16. An active cooling system for photovoltaic modules

    International Nuclear Information System (INIS)

    Teo, H.G.; Lee, P.S.; Hawlader, M.N.A.

    2012-01-01

    The electrical efficiency of photovoltaic (PV) cell is adversely affected by the significant increase of cell operating temperature during absorption of solar radiation. A hybrid photovoltaic/thermal (PV/T) solar system was designed, fabricated and experimentally investigated in this work. To actively cool the PV cells, a parallel array of ducts with inlet/outlet manifold designed for uniform airflow distribution was attached to the back of the PV panel. Experiments were performed with and without active cooling. A linear trend between the efficiency and temperature was found. Without active cooling, the temperature of the module was high and solar cells can only achieve an efficiency of 8–9%. However, when the module was operated under active cooling condition, the temperature dropped significantly leading to an increase in efficiency of solar cells to between 12% and 14%. A heat transfer simulation model was developed to compare to the actual temperature profile of PV module and good agreement between the simulation and experimental results is obtained.

  17. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

  18. Emergency cooling system for the PHENIX reactor

    International Nuclear Information System (INIS)

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

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

  19. Two-phase flow experiments in emergency core cooling feed through the hot leg for developing numerical models

    International Nuclear Information System (INIS)

    Staebler, T.; Meyer, L.; Schulenberg, T.; Laurien, E.

    2006-01-01

    When a leakage, a 'loss-of-coolant accident', occurs in a light water reactor, the emergency cooling system is able to supply large amounts of coolant to ensure residual heat removal. This supply can be routed through a special emergency cooling pipe, the 'scoop', into the horizontal section of the main coolant pipe, the 'hot leg'. At the same time, hot steam from the superheated, partly voided core flows against the coolant. This gives rise to a two-phase flow in the opposite direction. A factor of primary interest in this situation is whether the coolant supplied by the emergency cooling system will reach the reactor core. The research project is being conducted in order to compute the rate of water supply by numerical methods. The WENKA test facility has been designed and built at the Karlsruhe Research Center to verify numerical calculations. It can be used to study the fluid dynamics phenomena expected to arise in emergency coolant feeding into the hot leg; the necessary local data can be determined experimentally. An extensive database for validating the numerical calculations is then available to complete the experimental work. (orig.)

  20. Cooling system for auxiliary systems of a nuclear power plant

    International Nuclear Information System (INIS)

    Maerker, W.; Mueller, K.; Roller, W.

    1981-01-01

    From the reactor auxiliary and ancillary systems of a nuclear facility heat has to be removed without the hazard arising that radioactive liquids or gases may escape from the safe area of the nuclear facility. A cooling system is described allowing at every moment to make available cooling fluid at a temperature sufficiently low for heat exchangers to be able to remove the heat from such auxiliary systems without needing fresh water supply or water reservoirs. For this purpose a dry cooling tower is connected in series with a heat exchanger that is cooled on the secondary side by means of a refrigerating machine. The cooling pipes are filled with a nonfreezable fluid. By means of a bypass a minimum temperature is guaranteed at cold weather. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  2. New Sodium Cooled Long-Life Cores with Axially Multi-Driver Regions

    International Nuclear Information System (INIS)

    Hyun, Hae Ri; Hong, Ser Gi

    2014-01-01

    In this concept of long-life core (they are sometimes called B-B (Breed and Burn)), tall blanket is placed above the relatively short driver fuel. In the initial stage of burning, the power by fission is mostly generated in the driver region and it moves into the blanket region. The power and flux distributions that are highly peaked in the axial direction propagates slowly from the driver into the blanket region. This concept of long-life core fully utilizes the breeding of blanket in the fast spectra and it can achieve very high burnup of fuel. In this work, we introduce new sodium cooled longlife cores rating 600MWe (1800MWt). In these cores, the driver regions are heterogeneously placed into blanket region so as to achieve stabilized and less peaked axial power distribution as depletion proceeds. At present, our study is focused on only two axial driver regions but this concept can be easily extended onto the multi-driver region concept. The cores designed in this paper have two axial driver regions so as to have stabilized and less peaked axial power distributions as depletion proceeds. The results of the core design and analyses show that the cores have very long-lives longer than -49EFPYs and high discharge burnup higher than 200GWD/kg. Additionally, we considered a long-life core having no blanket. As expected, it was shown that these cores have stabilized and less peaked axial power distribution as the fuel depletes. However, the study shows that the cores having two driver regions still show high initial peaking of the axial power distributions and the core can be optimized by changing the driver fuel height

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

    Science.gov (United States)

    Marra, John J

    2014-04-01

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

  4. Controlled cooling of an electronic system based on projected conditions

    Science.gov (United States)

    David, Milnes P.; Iyengar, Madhusudan K.; Schmidt, Roger R.

    2015-08-18

    Energy efficient control of a cooling system cooling an electronic system is provided based, in part, on projected conditions. The control includes automatically determining an adjusted control setting(s) for an adjustable cooling component(s) of the cooling system. The automatically determining is based, at least in part, on projected power consumed by the electronic system at a future time and projected temperature at the future time of a heat sink to which heat extracted is rejected. The automatically determining operates to reduce power consumption of the cooling system and/or the electronic system while ensuring that at least one targeted temperature associated with the cooling system or the electronic system is within a desired range. The automatically determining may be based, at least in part, on an experimentally obtained model(s) relating the targeted temperature and power consumption of the adjustable cooling component(s) of the cooling system.

  5. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

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

    1987-05-01

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

  6. Analysis of passive moderator cooling system of Candu-6A reactor at emergency condition

    International Nuclear Information System (INIS)

    Umar, Efrizon; Subki, M. Hadid; Vecchiarelli, Jack

    2001-01-01

    Analysis of passive moderator cooling system subject to in-core LOCA with no emergency core cooling injection has been done. In this study, the new model of passive moderator system has been tested for emergency conditions and CATHENA code Mod-3.5b/Rev1 is used to calculate some parameters of this passive moderator cooling system. This result of simulation show that the proposed moderator cooling system have given satisfactory result, especially for the case with 0.7 m riser diameter and the number of heat exchanger tubes 8100. For PEWS tank containing 3000 m3 of light water initially at 30 0C and a 3641 m2 moderator heat exchanger, the average long-term heat removed rate balances the moderator heat load and the flow through the passive moderator loop remains stable for over 72 hours with no saturated boiling in the calandria and flow instabilities do not develop during long-term period

  7. Electromechanically cooled germanium radiation detector system

    International Nuclear Information System (INIS)

    Lavietes, Anthony D.; Joseph Mauger, G.; Anderson, Eric H.

    1999-01-01

    We have successfully developed and fielded an electromechanically cooled germanium radiation detector (EMC-HPGe) at Lawrence Livermore National Laboratory (LLNL). This detector system was designed to provide optimum energy resolution, long lifetime, and extremely reliable operation for unattended and portable applications. For most analytical applications, high purity germanium (HPGe) detectors are the standard detectors of choice, providing an unsurpassed combination of high energy resolution performance and exceptional detection efficiency. Logistical difficulties associated with providing the required liquid nitrogen (LN) for cooling is the primary reason that these systems are found mainly in laboratories. The EMC-HPGe detector system described in this paper successfully provides HPGe detector performance in a portable instrument that allows for isotopic analysis in the field. It incorporates a unique active vibration control system that allows the use of a Sunpower Stirling cycle cryocooler unit without significant spectral degradation from microphonics. All standard isotopic analysis codes, including MGA and MGA++, GAMANL, GRPANL and MGAU, typically used with HPGe detectors can be used with this system with excellent results. Several national and international Safeguards organisations including the International Atomic Energy Agency (IAEA) and U.S. Department of Energy (DOE) have expressed interest in this system. The detector was combined with custom software and demonstrated as a rapid Field Radiometric Identification System (FRIS) for the U.S. Customs Service . The European Communities' Safeguards Directorate (EURATOM) is field-testing the first Safeguards prototype in their applications. The EMC-HPGe detector system design, recent applications, and results will be highlighted

  8. Passive afterheat removal in the HTGR with the liner cooling system as a heat sink

    International Nuclear Information System (INIS)

    Rehm, W.; Jahn, W.; Verfondern, K.

    1984-09-01

    The report deals with the transients of temperature and system pressure and the fission product behaviour in the primary circuit of an HTGR during passive afterheat removal, where the liner cooling system of the PCRV serves as a heat sink. The analysis has been made for the PNP-500-reactor representing nuclear plants with medium thermal power. The investigations show that the liner cooling system is able to control a core heatup. High temperature loads are encountered in the upper core region. In the case of a reactor under pressure the fuel elements and the primary circuit remain intact as the first and second barriers for fission products. In the case of a depressurized primary circuit the liner cooling system also keeps the PCRV at normal operating temperatures. The effects of a core heatup on component damage and release of fission products are thus limited. (orig.) [de

  9. Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Hsu, C.J.

    1983-01-01

    In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

  10. Cooling Tower (Evaporative Cooling System) Measurement and Verification Protocol

    Energy Technology Data Exchange (ETDEWEB)

    Kurnik, Charles W. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Boyd, Brian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Stoughton, Kate M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lewis, Taylor [Colorado Energy Office, Denver, CO (United States)

    2017-12-05

    This measurement and verification (M and V) protocol provides procedures for energy service companies (ESCOs) and water efficiency service companies (WESCOs) to determine water savings resulting from water conservation measures (WCMs) in energy performance contracts associated with cooling tower efficiency projects. The water savings are determined by comparing the baseline water use to the water use after the WCM has been implemented. This protocol outlines the basic structure of the M and V plan, and details the procedures to use to determine water savings.

  11. THE RELATION BETWEEN COOL CLUSTER CORES AND HERSCHEL-DETECTED STAR FORMATION IN BRIGHTEST CLUSTER GALAXIES

    Energy Technology Data Exchange (ETDEWEB)

    Rawle, T. D.; Egami, E.; Rex, M.; Fiedler, A.; Haines, C. P.; Pereira, M. J.; Portouw, J.; Walth, G. [Steward Observatory, University of Arizona, 933 N. Cherry Ave., Tucson, AZ 85721 (United States); Edge, A. C. [Institute for Computational Cosmology, Durham University, South Road, Durham DH1 3LE (United Kingdom); Smith, G. P. [School of Physics and Astronomy, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); Altieri, B.; Valtchanov, I. [Herschel Science Centre, ESAC, ESA, P.O. Box 78, Villanueva de la Canada, 28691 Madrid (Spain); Perez-Gonzalez, P. G. [Departamento de Astrofisica, Facultad de CC. Fisicas, Universidad Complutense de Madrid, E-28040 Madrid (Spain); Van der Werf, P. P. [Sterrewacht Leiden, Leiden University, P.O. Box 9513, 2300 RA, Leiden (Netherlands); Zemcov, M., E-mail: trawle@as.arizona.edu [Department of Physics, Mathematics and Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States)

    2012-03-01

    We present far-infrared (FIR) analysis of 68 brightest cluster galaxies (BCGs) at 0.08 < z < 1.0. Deriving total infrared luminosities directly from Spitzer and Herschel photometry spanning the peak of the dust component (24-500 {mu}m), we calculate the obscured star formation rate (SFR). 22{sup +6.2}{sub -5.3}% of the BCGs are detected in the far-infrared, with SFR = 1-150 M{sub Sun} yr{sup -1}. The infrared luminosity is highly correlated with cluster X-ray gas cooling times for cool-core clusters (gas cooling time <1 Gyr), strongly suggesting that the star formation in these BCGs is influenced by the cluster-scale cooling process. The occurrence of the molecular gas tracing H{alpha} emission is also correlated with obscured star formation. For all but the most luminous BCGs (L{sub TIR} > 2 Multiplication-Sign 10{sup 11} L{sub Sun }), only a small ({approx}<0.4 mag) reddening correction is required for SFR(H{alpha}) to agree with SFR{sub FIR}. The relatively low H{alpha} 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 from normal stellar mass loss.

  12. The Role of Cerenkov Radiation in the Pressure Balance of Cool Core Clusters of Galaxies

    Energy Technology Data Exchange (ETDEWEB)

    Lieu, Richard [Department of Physics, University of Alabama, Huntsville, AL 35899 (United States)

    2017-03-20

    Despite the substantial progress made recently in understanding the role of AGN feedback and associated non-thermal effects, the precise mechanism that prevents the core of some clusters of galaxies from collapsing catastrophically by radiative cooling remains unidentified. In this Letter, we demonstrate that the evolution of a cluster's cooling core, in terms of its density, temperature, and magnetic field strength, inevitably enables the plasma electrons there to quickly become Cerenkov loss dominated, with emission at the radio frequency of ≲350 Hz, and with a rate considerably exceeding free–free continuum and line emission. However, the same does not apply to the plasmas at the cluster's outskirts, which lacks such radiation. Owing to its low frequency, the radiation cannot escape, but because over the relevant scale size of a Cerenkov wavelength the energy of an electron in the gas cannot follow the Boltzmann distribution to the requisite precision to ensure reabsorption always occurs faster than stimulated emission, the emitting gas cools before it reheats. This leaves behind the radiation itself, trapped by the overlying reflective plasma, yet providing enough pressure to maintain quasi-hydrostatic equilibrium. The mass condensation then happens by Rayleigh–Taylor instability, at a rate determined by the outermost radius where Cerenkov radiation can occur. In this way, it is possible to estimate the rate at ≈2 M {sub ⊙} year{sup −1}, consistent with observational inference. Thus, the process appears to provide a natural solution to the longstanding problem of “cooling flow” in clusters; at least it offers another line of defense against cooling and collapse should gas heating by AGN feedback be inadequate in some clusters.

  13. Improvements in or relating to cooling systems for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Ljubivy, A.G.; Batjukov, V.I.; Shkhian, T.G.; Fadeev, A.I.

    1980-01-01

    A cooling system is proposed which can be used to cool a set of nuclear fuel assemblies arranged in a reactor core or placed in a container for spent fuel assemblies. The object of the invention is to provide a system which would prevent leakage of coolant from the vessel in the event of a rupture of the coolant supply pipeline externally of the vessel. In the case of the reactor cooling system the level of the coolant is stopped from dropping below the level of the active portion of the fuel assemblies and thus prevents a breakdown of the reactor. (UK)

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

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Hori, Toru; Konomura, Mamoru

    2003-07-01

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

  15. Neutronic design for a 100MW{sub th} Small modular natural circulation lead or lead-alloy cooled fast reactors core

    Energy Technology Data Exchange (ETDEWEB)

    Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q., E-mail: shchshch@ustc.edu.cn, E-mail: hlchen1@ustc.edu.cn, E-mail: kulah@mail.ustc.edu.cn, E-mail: zchen214@mail.ustc.edu.cn, E-mail: zengqin@ustc.edu.cn [Univ. of Science and Technology of China, School of Nuclear Science and Technology, Hefei, Anhui (China)

    2015-07-01

    Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW{sub th} natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

  16. Estimation on the Pressure Loss of the Conceptual Primary Cooling System and Design of the Primary Cooling Pump for a Research Reactor

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Oh, Jae Min; Park, Jong Hark; Chae, Hee Taek; Seo, Jae Kwang; Park, Cheon Tae; Yoon, Ju Hyeon; Lee, Doo Jeong

    2009-01-01

    A new conceptual primary cooling system (PCS) for a research reactor has been designed for an adequate cooling to the reactor core which has various powers ranging from 30MW through 80MW. The developed primary cooling system consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. Because the system flow rate should be determined by the thermal hydraulic design analysis for the core, the heads to design the primary cooling pumps (PCPs) in a PCS will be estimated by the variable system flow rates. The heads of the part of a research reactor vessel was evaluated by the previous study. The various pressure losses of the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. The purpose of this research is to estimate the various pressure losses and to design the PCPs

  17. Cooling system for the IFMIF-EVEDA radiofrequency system

    International Nuclear Information System (INIS)

    Perez Pichel, G. D.

    2012-01-01

    The IFMIF-EVEDA project consists on an accelerator prototype that will be installed at Rokkasho (Japan). Through CIEMAT, that is responsible of the development of many systems and components. Empresarios Agrupados get the responsibility of the detailed design of the cooling system for the radiofrequency system (RF system) that must feed the accelerator. the RF water cooling systems is the water primary circuit that provides the required water flow (with a certain temperature, pressure and water quality) and also dissipates the necessary thermal power of all the radiofrequency system equipment. (Author) 4 refs.

  18. Development of the interactive model between Component Cooling Water System and Containment Cooling System using GOTHIC

    International Nuclear Information System (INIS)

    Byun, Choong Sup; Song, Dong Soo; Jun, Hwang Yong

    2006-01-01

    In a design point of view, component cooling water (CCW) system is not full-interactively designed with its heat loads. Heat loads are calculated from the CCW design flow and temperature condition which is determined with conservatism. Then the CCW heat exchanger is sized by using total maximized heat loads from above calculation. This approach does not give the optimized performance results and the exact trends of CCW system and the loads during transient. Therefore a combined model for performance analysis of containment and the component cooling water(CCW) system is developed by using GOTHIC software code. The model is verified by using the design parameters of component cooling water heat exchanger and the heat loads during the recirculation mode of loss of coolant accident scenario. This model may be used for calculating the realistic containment response and CCW performance, and increasing the ultimate heat sink temperature limits

  19. 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

  20. Safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release

    International Nuclear Information System (INIS)

    Pointner, W.; Broecker, A.

    2012-01-01

    The report on safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release covers the following issues: assessment of the relevant status for PWR, evaluation of the national and international (USA, Canada, France) status, actualization of recommendations, transferability from PWR to BWR. Generic studies on the core cooling capability in case of insulation material release in BWR-type reactors were evaluated.

  1. Evaluation of heat exchange performance for the auxiliary component cooling water system cooling tower in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Kameyama, Yasuhiko; Shimizu, Atsushi; Inoi, Hiroyuki; Yamazaki, Kazunori; Shimizu, Yasunori; Aragaki, Etsushi; Ota, Yukimaru; Fujimoto, Nozomu

    2006-09-01

    The auxiliary component cooling water system (ACCWS) is one of the cooling system in High Temperature Engineering Test Reactor (HTTR). The ACCWS has main two features, many facilities cooling, and heat sink of the vessel cooling system which is one of the engineering safety features. Therefore, the ACCWS is required to satisfy the design criteria of heat removal performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the ACCWS cooling tower was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that ACCWS cooling tower has the required heat exchange performance in the design. (author)

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

    International Nuclear Information System (INIS)

    Ha, J.J.; Belhadj, M.; Aldemir, T.; Christensen, R.N.

    1987-01-01

    Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16 N 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)

  3. Performance Optimization of the Water Cooling System for Resonance Frequency Control of the PEFP DTL

    International Nuclear Information System (INIS)

    Kim, K. Y.; Kim, H. K.; Kim, H. S.; Yoon, J. C.; Sohn, Y. K.; Kweon, S. J.; Park, J.; Kim, K. S.

    2010-03-01

    The objective of in this research project is prototype cooling water skid of separated closed loop in order to supply and withdraw low conductivity deionized water in drift tube of drift tube linac as core components of proton accelerates. This report is dealt with design specification of J-PARC 400 MeV Linac cooling water system, PEFP DTL cooling system, specification of RCCS21-24, RCCS101 with pump, loss coefficient for DTL2 modeling, pressure drop with flow rate of heat exchanger.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  5. Misting-cooling systems for microclimatic control in public space

    OpenAIRE

    Nunes, Joao; Zoilo, Inaki; Jacinto, Nuno; Nunes, Ana; Torres-Campos, Tiago; Pacheco, Manuel; Fonseca, David

    2011-01-01

    Misting-cooling systems have been used in outdoor spaces mainly for aesthetic purposes, and punctual cooling achievement. However, they can be highly effective in outdoor spaces’ bioclimatic comfort, in terms of microclimatic control, as an evaporative cooling system. Recent concerns in increasing bioclimatic standards in public outdoor spaces, along with more sustainable practices, gave origin to reasoning where plastic principles are combined with the study of cooling efficacy, in order to ...

  6. Investigations of combined used of cooling ponds with cooling towers or spraying systems

    International Nuclear Information System (INIS)

    Farforovsky, V.B.

    1990-01-01

    Based on a brief analysis of the methods of investigating cooling ponds, spraying systems and cooling towers, a conclusion is made that the direct modelling of the combined use of cooling systems listed cannot be realized. An approach to scale modelling of cooling ponds is proposed enabling all problems posed by the combined use of coolers to be solved. Emphasized is the importance of a proper choice of a scheme of including a cooler in a general water circulation system of thermal and nuclear power plants. A sequence of selecting a cooling tower of the type and spraying system of the size ensuring the specified temperature regime in a water circulation system is exemplified by the water system of the Ghorasal thermal power plant in Bangladesh

  7. Emergency cooling system for nuclear reactors

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

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

  8. Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

    International Nuclear Information System (INIS)

    Ginsberg, T.; Chen, J.C.

    1984-01-01

    Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

  9. Laser-cooled atoms inside a hollow-core photonic-crystal fiber

    DEFF Research Database (Denmark)

    Bajcsy, Michal; Hofferberth, S.; Peyronel, Thibault

    2011-01-01

    We describe the loading of laser-cooled rubidium atoms into a single-mode hollow-core photonic-crystal fiber. Inside the fiber, the atoms are confined by a far-detuned optical trap and probed by a weak resonant beam. We describe different loading methods and compare their trade-offs in terms...... of implementation complexity and atom-loading efficiency. The most efficient procedure results in loading of ∼30,000 rubidium atoms, which creates a medium with an optical depth of ∼180 inside the fiber. Compared to our earlier study this represents a sixfold increase in the maximum achieved optical depth...

  10. Reactor core design aiding system

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Hamaguchi, Yukio; Nakao, Takashi; Kondo, Yasuhide

    1995-01-01

    A two-dimensional radial power distribution and an axial one-dimensional power distribution are determined based on a distribution of a three-dimensional infinite multiplication factor, to obtain estimated power distribution estimation values. The estimation values are synthesized to obtain estimated three-dimensional power distribution values. In addition, the distribution of a two-dimensional radial multiplication factor and the distribution of an one-dimensional axial multiplication factor are determined based on the three-dimensional power distribution, to obtain estimated values for the multiplication factor distribution. The estimated values are synthesized to form estimated values for the three-dimensional multiplication factor distribution. Further, estimated fuel loading pattern value is determined based on the three-dimensional power distribution or the two-dimensional radial power distribution. Since the processes for determining the estimated values comprise only additive and multiplying operations, processing time can be remarkably saved compared with calculation based on a detailed physical models. Since the estimation is performed on every fuel assemblies, a nervous circuit network not depending on the reactor core system can be constituted. (N.H.)

  11. Performance characteristic of hybrid cooling system based on cooling pad and evaporator

    Science.gov (United States)

    Yoon, J. I.; Son, C. H.; Choi, K. H.; Kim, Y. B.; Sung, Y. H.; Roh, S. J.; Kim, Y. M.; Seol, S. H.

    2018-01-01

    In South Korea, most of domestic animals such as pigs and chickens might die due to thermal diseases if they are exposed to the high temperature consistently. In order to save them from the heat wave, numerous efforts have been carried out: installing a shade net, adjusting time of feeding, spraying mist and setting up a circulation fan. However, these methods have not shown significant improvements. Thus, this study proposes a hybrid cooling system combining evaporative cooler and air-conditioner in order to resolve the conventional problems caused by the high temperature in the livestock industry. The problem of cooling systems using evaporative cooling pads is that they are not effective for eliminating huge heat load due to their limited capacity. And, temperature of the supplied air cannot be low enough compared to conventional air-conditioning systems. On the other hand, conventional air-conditioning systems require relatively expensive installation cost, and high operating cost compared to evaporative cooling system. The hybrid cooling system makes up for the lack of cooling capacity of the evaporative cooler by employing the conventional air-conditioner. Additionally, temperature of supplied air can be lowered enough. In the hybrid cooling system, induced air by a fan is cooled by the evaporation of water in the cooling pad, and it is cooled again by an evaporator in the air-conditioner. Therefore, the more economical operation is possible due to additionally obtained cooling capacity from the cooling pads. Major results of experimental analysis of hybrid cooling system are as follows. The compressor power consumption of the hybrid cooling system is about 23% lower, and its COP is 17% higher than that of the conventional air-conditioners. Regarding the condition of changing ambient temperature, the total power consumption decreased by about 5% as the ambient temperature changed from 28.7°C to 31.7°C. Cooling capacity and COP also presented about 3% and 1

  12. The cooling history and the depth of detachment faulting at the Atlantis Massif oceanic core complex

    Science.gov (United States)

    Schoolmeesters, Nicole; Cheadle, Michael J.; John, Barbara E.; Reiners, Peter W.; Gee, Jeffrey; Grimes, Craig B.

    2012-10-01

    Oceanic core complexes (OCCs) are domal exposures of oceanic crust and mantle interpreted to be denuded to the seafloor by large slip oceanic detachment faults. We combine previously reported U-Pb zircon crystallization ages with (U-Th)/He zircon thermochronometry and multicomponent magnetic remanence data to determine the cooling history of the footwall to the Atlantis Massif OCC (30°N, MAR) and help establish cooling rates, as well as depths of detachment faulting and gabbro emplacement. We present nine new (U-Th)/He zircon ages for samples from IODP Hole U1309D ranging from 40 to 1415 m below seafloor. These data paired with U-Pb zircon ages and magnetic remanence data constrain cooling rates of gabbroic rocks from the upper 800 m of the central dome at Atlantis Massif as 2895 (+1276/-1162) °C Myr-1 (from ˜780°C to ˜250°C); the lower 600 m of the borehole cooled more slowly at mean rates of ˜500 (+125/-102) °C Myr-1(from ˜780°C to present-day temperatures). Rocks from the uppermost part of the hole also reveal a brief period of slow cooling at rates of ˜300°C Myr-1, possibly due to hydrothermal circulation to ˜4 km depth through the detachment fault zone. Assuming a fault slip rate of 20 mm/yr (from U-Pb zircon ages of surface samples) and a rolling hinge model for the sub-surface fault geometry, we predict that the 780°C isotherm lies at ˜7 km below the axial valley floor, likely corresponding both to the depth at which the semi-brittle detachment fault roots and the probable upper limit of significant gabbro emplacement.

  13. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  14. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  15. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-09

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

  16. Seismic response of high temperature gas-cooled reactor core with block-type fuel, (2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi; Honma, Toshiaki.

    1980-01-01

    For the aseismic design of a high temperature gas-cooled reactor (HTGR) with block-type fuel, it is necessary to predict the motion and force of core columns and blocks. To reveal column vibration characteristics in three-dimensional space and impact response, column vibration tests were carried out with a scale model of a one-region section (seven columns) of the HTGR core. The results are as follows: (1) the column has a soft spring characteristic based on stacked blocks connected with loose pins, (2) the column has whirling phenomena, (3) the compression spring force simulating the gas pressure has the effect of raising the column resonance frequency, and (4) the vibration behavior of the stacked block column and impact response of the surrounding columns show agreement between experiment and analysis. (author)

  17. Modelling guidelines for core exit temperature simulations with system codes

    Energy Technology Data Exchange (ETDEWEB)

    Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

    2015-05-15

    Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

  18. Phasing of Debuncher Stochastic Cooling Transverse Systems

    International Nuclear Information System (INIS)

    Pasquinelli, Ralph

    2000-01-01

    With the higher frequency of the cooling systems in the Debuncher, a modified method of making transfer functions has been developed for transverse systems. (Measuring of the momentum systems is unchanged.) Speed in making the measurements is critical, as the beam tends to decelerate due to vacuum lifetime. In the 4-8 GHz band, the harmonics in the Debuncher are 6,700 to 13,400 times the revolution frequency. Every Hertz change in revolution frequency is multiplied by this harmonic number and becomes a frequency measurement error, which is an appreciable percent of the momentum width of the beam. It was originally thought that a momentum cooling system would be phased first so that the beam could be kept from drifting in revolution frequency. As it turned out, the momentum cooling was so effective (even with the gain turned down) that the momentum width normalized to fo became less than one Hertz on the Schottky pickup. A beam this narrow requires very precise measurement of tune and revolution frequency. It was difficult to get repeatable results. For initial measuring of the transverse arrays, relative phase and delay is all that is required, so the measurement settings outlined below will suffice. Once all input and output arrays are phased, a more precise measurement of all pickups to all kickers can be done with more points and both upper and lower side bands, as in figure 1. Settings on the network analyzer were adjusted for maximum measurement speed. Data is not analyzed until a complete set of measurements is taken. Start and stop frequencies should be chosen to be just slightly wider than the band being measured. For transverse systems, select betatron USB for the measurement type. This will make the measurement two times faster. Select 101 for the number of points, sweep time of 5 seconds, IF bandwidth 30 Hz, averages = 1. It is important during the phasing to continually measure the revolution frequency and beam width of the beam for transverse systems

  19. Deposit control in process cooling water systems

    International Nuclear Information System (INIS)

    Venkataramani, B.

    1981-01-01

    In order to achieve efficient heat transfer in cooling water systems, it is essential to control the fouling of heat exchanger surfaces. Solubilities of scale forming salts, their growth into crystals, and the nature of the surfaces play important roles in the deposition phenomenon. Condensed phosphates, organic polymers and compounds like phosphates are effective in controlling deposition of scale forming salts. The surface active agents inhibit crystal growth and modify the crystals of the scale forming salts, and thus prevent deposition of dense, uniformly structured crystalline mass on the heat transfer surface. Understanding the mechanism of biofouling is essential to control it by surface active agents. Certain measures taken in the plant, such as back flushing, to control scaling, sometimes may not be effective and can be detrimental to the system itself. (author)

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

    International Nuclear Information System (INIS)

    Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

    2015-01-01

    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

  1. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myoung Jun; Lee, Hee Joon [Kookmin Univ., Seoul (Korea, Republic of); Moon, Joo Hyung; Bae, Youngmin; Kim, Youngin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

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

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

    International Nuclear Information System (INIS)

    Kim, Myoung Jun; Lee, Hee Joon; Moon, Joo Hyung; Bae, Youngmin; Kim, Youngin

    2014-01-01

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

  4. CoolPack – Simulation tools for refrigeration systems

    DEFF Research Database (Denmark)

    Jakobsen, Arne; Rasmussen, Bjarne D.; Andersen, Simon Engedal

    1999-01-01

    CoolPack is a collection of programs used for energy analysis and optimisation of refrigeration systems. CoolPack is developed at the Department of Energy Engineering at the Technical University of Denmark. The Danish Energy Agency finances the project. CoolPack is freeware and can be downloaded...

  5. Renewal of cooling system of JMTR

    International Nuclear Information System (INIS)

    Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Koike, Sumio; Nishiyama, Yutaka; Fukasaku, Akitomi

    2011-06-01

    The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. The refurbishment work was started from 2007, and restart is planned in 2011. Renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities whose replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replacement priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

  6. Core sampling system spare parts assessment

    International Nuclear Information System (INIS)

    Walter, E.J.

    1995-01-01

    Soon, there will be 4 independent core sampling systems obtaining samples from the underground tanks. It is desirable that these systems be available for sampling during the next 2 years. This assessment was prepared to evaluate the adequacy of the spare parts identified for the core sampling system and to provide recommendations that may remediate overages or inadequacies of spare parts

  7. Core Science Systems--Mission overview

    Science.gov (United States)

    Gallagher, Kevin T.

    2012-01-01

    The Core Science Systems Mission Area delivers nationally focused Earth systems and information science that provides fundamental research and data that underpins all Mission Areas of the USGS, the USGS Science Strategy, and Presidential, Secretarial, and societal priorities. —Kevin T. Gallagher, Associate Director, Core Science Systems

  8. Modelling and analysis of a desiccant cooling system using the regenerative indirect evaporative cooling process

    DEFF Research Database (Denmark)

    Bellemo, Lorenzo; Elmegaard, Brian; Reinholdt, Lars O.

    2013-01-01

    This paper focuses on the numerical modeling and analysis of a Desiccant Cooling (DEC) system with regenerative indirect evaporative cooling, termed Desiccant Dewpoint Cooling (DDC) system. The DDC system includes a Desiccant Wheel (DW), Dew Point Coolers (DPCs), a heat recovery unit and a heat...... in different climates: temperate in Copenhagen and Mediterranean in Venice. Cheap and clean heat sources (e.g. solar energy) strongly increase the attractiveness of the DDC system. For the Mediterranean climate the DDC system represents a convenient alternative to chiller-based systems in terms of energy costs...... and CO2 emissions. The electricity consumption for auxiliaries in the DDC system is higher than in the chiller-based systems. The number of commercial-size DPC units required to cover the cooling load during the whole period is high: 8 in Copenhagen and 12 in Venice....

  9. Installation of JMTR core management system

    International Nuclear Information System (INIS)

    Imaizumi, Tomomi; Ide, Hiroshi; Naka, Michihiro; Komukai, Bunsaku; Nagao, Yoshiharu

    2013-01-01

    In order to carry out the core management after the reoperation of JMTR quickly and accurately, the authors took up the Standard Reactor Analysis Code (SRAC) system and core management support programs that are operating in a general-purpose large computer and transferred them to PC (OS: Linux), and newly established a JMTR core management system. As for the core analysis, this measure enabled an increase in the processing speed from the check of core arrangement to the result display of nuclear restriction values to about 60 times, compared with the conventional method. It was confirmed that the differences of calculation results originated from the difference of internal display of computers, associated with the transfer of each analysis code from GS21-400 system to PC-Linux, were within practically allowable level. In the future, this system will be applied to the core analysis of JMTR, as well as to the preparation of operation plans. (A.O.)

  10. Conceptual core design study for Japan sodium-cooled fast reactor: Review of sodium void reactivity worth evaluation

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2012-01-01

    The conceptual core design study for a large-scale Japan sodium-cooled fast reactor (JSFR) have been carried out in the framework of the FaCT project. The reference “High-internal conversion” core can satisfy the requirements for enhanced safety, as well as achieving economic competitiveness. In order to increase the design reliability, more rigorous uncertainty evaluation is important. Development of the verification and validation methodology of the core neutronic design method is currently underway. (author)

  11. SNS Resonance Control Cooling Systems and Quadrupole Magnet Cooling Systems DIW Chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Magda, Karoly [ORNL

    2018-01-01

    This report focuses on control of the water chemistry for the Spallation Neutron Source (SNS) Resonance Control Cooling System (RCCS)/Quadrupole Magnet Cooling System (QMCS) deionized water (DIW) cooling loops. Data collected from spring 2013 through spring 2016 are discussed, and an operations regime is recommended.It was found that the RCCS operates with an average pH of 7.24 for all lines (from 7.0 to 7.5, slightly alkaline), the average low dissolved oxygen is in the area of < 36 ppb, and the main loop average resistivity of is > 14 MΩ-cm. The QMCS was found to be operating in a similar regime, with a slightly alkaline pH of 7.5 , low dissolved oxygen in the area of < 45 ppb, and main loop resistivity of 10 to 15 MΩ-cm. During data reading, operational corrections were done on the polishing loops to improve the water chemistry regime. Therefore some trends changed over time.It is recommended that the cooling loops operate in a regime in which the water has a resistivity that is as high as achievable, a dissolved oxygen concentration that is as low as achievable, and a neutral or slightly alkaline pH.

  12. Nuclear reactor equipped with a flooding tank and a residual heat removal and emergency cooling system

    International Nuclear Information System (INIS)

    Schabert, H.P.; Winkler, F.

    1975-01-01

    A description is given of a nuclear reactor such as a pressurized-water reactor or the like which is equipped with a flooding tank and a residual heat removal and emergency cooling system. The flooding tank is arranged within the containment shell at an elevation above the upper edge of the reactor core and contains a liquid for flooding the reactor core in the event of a loss of coolant

  13. 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.

  14. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

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

  15. PEP cooling water systems and underground piped utilities design criteria report

    International Nuclear Information System (INIS)

    Hall, F.; Robbins, D.

    1975-10-01

    This paper discusses the cooling systems required by the PEP Storage Ring. Particular topics discussed are: Cooling tower systems, RF cavity and vacuum chamber LCW cooling systems, klystron and ring magnet LLW cooling systems, Injection magnet LCW Cooling Systems; PEP interaction area detector LCW Cooling Systems; and underground piped utilities. 1 ref., 20 figs

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    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)

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

    International Nuclear Information System (INIS)

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    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)

  19. Cooling systems research at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Spigarelli, S.A.

    1977-01-01

    Studies of the thermal plumes resulting from discharges from once-through cooling systems of electric generating stations are reviewed. The collection of large amounts of water temperature data for definition of the three-dimensional structure of a thermal plume, of current data, and related ambient data for model evaluation purposes required the development of an integrated data collection system. The Argonne system employs measurements of water temperature over the water column from a moving small boat. Temperatures are measured with thermistors attached to a rigid strut for surface plumes and to a flexible, faired cable for submerged plumes. Water temperatures and boat location, determined by a microwave ranging system, are recorded on magnetic tape while the boat is underway and prove a quasi-synoptic map of plume temperatures. Automated data handling and processing procedures provide for the production of isotherm maps of the plume at several elevations and in cross section. Mathematical model evaluation for surface discharges of waste heat included the consideration of over 40 different models and detailed evaluation of 11 models. Most models were run on Argonne's computers, and all models were evaluated in terms of their limitations and capabilities as well as their predictive performance against prototype data. Measurements were made of thermal plumes at the discharges of nuclear power plants located on the shores of Lake Michigan

  20. Simulation of an adsorption solar cooling system

    International Nuclear Information System (INIS)

    Hassan, H.Z.; Mohamad, A.A.; Bennacer, R.

    2011-01-01

    A more realistic theoretical simulation model for a tubular solar adsorption refrigerating system using activated carbon-methanol (AC/M) pair has been introduced. The mathematical model represents the heat and mass transfer inside the adsorption bed, the condenser, and the evaporator. The simulation technique takes into account the variations of ambient temperature and solar radiation along the day. Furthermore, the local pressure, and local thermal conductivity variations in space and time inside the tubular reactor are investigated as well. A C++ computer program is written to solve the proposed numerical model using the finite difference method. The developed program covers the operations of all the system components along the cycle time. The performance of the tubular reactor, the condenser, and the evaporator has been discussed. Time allocation chart and switching operations for the solar refrigeration system processes are illustrated as well. The case studied has a 1 m 2 surface area solar flat plate collector integrated with a 20 stainless steel tubes containing the AC/M pair and each tube has a 5 cm outer diameter. In addition, the condenser pressure is set to 54.2 kpa. It has been found that, the solar coefficient of performance and the specific cooling power of the system are 0.211 and 2.326 respectively. In addition, the pressure distribution inside the adsorption bed has been found nearly uniform and varying only with time. Furthermore, the AC/M thermal conductivity is shown to be constant in both space and time.

  1. Preliminary design concept of HYPER cooling system using Pb-Bi coolant

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam Il; Song, Tae Y.; Park, Won S.; Kim, Chang H

    2001-09-01

    The present study focuses on providing the basic concept of HYPER's cooling system based on simple and fundamental calculations. The system operating temperature was preliminarily determined as 340/510 .deg. C. The total system flow rate of HYPER is {approx} 40,000kg/sec and the flow velocity in the core is preliminarily designed to be {approx}1.5 m/sec. For hot conditions of HYPER core, the simple analytic calculation predicted that the maximum temperature of the cladding outer surface is 634 .deg. C, which is below the design limit, 650 .deg. C. However, the SLTHEN code modified for HYPER's subchannel analysis predicted that the maximum temperature of the cladding outer surface in the same conditions is higher than the design limit by 4.7 .deg. C. The comparison with the results of the analytic model and additional sensitivity calculations showed that the modified SLTHEN code can reasonably simulate the heat transfer between subchannels of the HYPER core and be used effectively for thermal hydraulic design of the HYPER core in conceptual design stage. A forced circulation is inevitable during a full power condition since natural circulation is not sufficient to cool the core with reasonable system pressure drop and reasonable system height. However, a natural circulation can be an excellent method for decay heat removal when the height difference between the core and the heat exchanger is above 10 m. In order to avoid high pressure loads on the vessel, loop configuration was chosen. The simplification of cooling system and high system efficiency were attained by removing independent target cooling system and intermediate heat transport system. A superheated rankle cycle was chosen since it is technically matured and its thermal efficiency is reasonably high.

  2. Reactor core operation management system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Tomomi.

    1992-05-28

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.).

  3. Reactor core operation management system

    International Nuclear Information System (INIS)

    Sato, Tomomi.

    1992-01-01

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.)

  4. EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Saito, Shinzo

    1977-09-01

    In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

  5. GOTHIC Simulation of Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Ha, Huiun; Kim, Hangon

    2013-01-01

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR

  6. GOTHIC Simulation of Passive Containment Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Kim, Hangon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR.

  7. Load calculations of radiant cooling systems for sizing the plant

    DEFF Research Database (Denmark)

    Bourdakis, Eleftherios; Kazanci, Ongun Berk; Olesen, Bjarne W.

    2015-01-01

    The aim of this study was, by using a building simulation software, to prove that a radiant cooling system should not be sized based on the maximum cooling load but at a lower value. For that reason six radiant cooling models were simulated with two control principles using 100%, 70% and 50......% of the maximum cooling load. It was concluded that all tested systems were able to provide an acceptable thermal environment even when the 50% of the maximum cooling load was used. From all the simulated systems the one that performed the best under both control principles was the ESCS ceiling system. Finally...... it was proved that ventilation systems should be sized based on the maximum cooling load....

  8. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    Science.gov (United States)

    Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  9. Integrated cooling system for the Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Johnson, B.; Chang, Y.

    1979-01-01

    The MFTF components that require water cooling include the neutral beam dumps, ion dumps, plasma dumps, baffle plates, magnet liners, gas boxes, streaming guns, and the neutral beam injectors. A total heat load of nearly 500 MW for 0.5 s dissipates over 4-min intervals. A steady-flow, closed-loop system is utilized. The design of the cooling system assumes that all components require cooling simultaneously. The cooling system contains process instrumentation for loop control. Alarms and safety interlocks are incorporated for the safe operation of the system

  10. Computer Aided Design of The Cooling System for Plastic Injection Molds

    Directory of Open Access Journals (Sweden)

    Hakan GÜRÜN

    2009-02-01

    Full Text Available The design of plastic injection molds and their cooling systems affect both the dimension, the shape, the quality of a plastic part and the cycle time of process and the cost of mold. In this study, the solid model design of a plastic injection mold and the design of cooling sysytem were possibly carried out without the designer interaction. Developed program permited the use of three types of the cooling system and the different cavity orientations and the multible plastic part placement into the mold cores. The program which was developed by using Visual LISP language and the VBA (Visual BASIC for Application modules, was applicated in the AutoCAD software domain. Trial studies were presented that the solid model design of plastic injection molds and the cooling systems increased the reliability, the flexibility and the speed of the design.

  11. Modern cooling systems in thermal power plants relieve environmental pollution. Pt. 2

    International Nuclear Information System (INIS)

    Brosche, D.

    1983-01-01

    Direct and indirect dry recirculation cooling, wet cooling tower, natural-draught wet cooling tower, combined cooling processes, hybrid cooling systems, cell cooling systems, auxiliary water preparation, cooling process design, afterheat removal in nuclear power plants, environmental effects, visible plumes as a function of weather conditions, environmental protection and energy supply assurance. (orig.) [de

  12. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

  13. Electric drive systems including smoothing capacitor cooling devices and systems

    Energy Technology Data Exchange (ETDEWEB)

    Dede, Ercan Mehmet; Zhou, Feng

    2017-02-28

    An electric drive system includes a smoothing capacitor including at least one terminal, a bus bar electrically coupled to the at least one terminal, a thermoelectric device including a first side and a second side positioned opposite the first side, where the first side is thermally coupled to at least one of the at least one terminal and the bus bar, and a cooling element thermally coupled to the second side of the thermoelectric device, where the cooling element dissipates heat from the thermoelectric device.

  14. The Role of Absorption Cooling for Reaching Sustainable Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Lindmark, Susanne

    2005-07-01

    This thesis focuses on the role and potential of absorption cooling in future energy systems. Two types of energy systems are investigated: a district energy system based on waste incineration and a distributed energy system with natural gas as fuel. In both cases, low temperature waste heat is used as driving energy for the absorption cooling. The main focus is to evaluate the absorption technology in an environmental perspective, in terms of reduced CO{sub 2} emissions. Economic evaluations are also performed. The reduced electricity when using absorption cooling instead of compression cooling is quantified and expressed as an increased net electrical yield. The results show that absorption cooling is an environmentally friendly way to produce cooling as it reduces the use of electrically driven cooling in the energy system and therefore also reduces global CO{sub 2} emissions. In the small-scale trigeneration system the electricity use is lowered with 84 % as compared to cooling production with compression chillers only. The CO{sub 2} emissions can be lowered to 45 CO{sub 2}/MWh{sub c} by using recoverable waste heat as driving heat for absorption chillers. However, the most cost effective cooling solution in a district energy system is a combination between absorption and compression cooling technologies according to the study. Absorption chillers have the potential to be suitable bottoming cycles for power production in distributed systems. Net electrical yields over 55 % may be reached in some cases with gas motors and absorption chillers. This small-scale system for cogeneration of power and cooling shows electrical efficiencies comparable to large-scale power plants and may contribute to reducing peak electricity demand associated with the cooling demand.

  15. Cryogenic system with the sub-cooled liquid nitrogen for cooling HTS power cable

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Y.F. [Chinese Academy of Sciences, Beijing (China). Technical Institute of Physics and Chemistry; Graduate School of Chinese Academy of Sciences, Beijing (China); Gong, L.H.; Xu, X.D.; Li, L.F.; Zhang, L. [Chinese Academy of Sciences, Beijing (China). Technical Institute of Physics and Chemistry; Xiao, L.Y. [Chinese Academy of Sciences, Beijing (China). Institute of Electrical Engineering

    2005-04-01

    A 10 m long, three-phase AC high-temperature superconducting (HTS) power cable had been fabricated and tested in China August 2003. The sub-cooled liquid nitrogen (LN{sub 2}) was used to cool the HTS cable. The sub-cooled LN{sub 2} circulation was built by means of a centrifugal pump through a heat exchanger in the sub-cooler, the three-phase HTS cable cryostats and a LN{sub 2} gas-liquid separator. The LN{sub 2} was cooled down to 65 K by means of decompressing, and the maximum cooling capacity was about 3.3 kW and the amount of consumed LN{sub 2} was about 72 L/h at 1500 A. Cryogenic system design, test and some experimental results would be presented in this paper. (author)

  16. ELECTRONIC CIRCUIT BOARDS NON-UNIFORM COOLING SYSTEM MODEL

    Directory of Open Access Journals (Sweden)

    D. V. Yevdulov

    2016-01-01

    Full Text Available Abstract. The paper considers a mathematical model of non-uniform cooling of electronic circuit boards. The block diagram of the system implementing this approach, the method of calculation of the electronic board temperature field, as well as the principle of its thermal performance optimizing are presented. In the considered scheme the main heat elimination from electronic board is produced by the radiator system, and additional cooling of the most temperature-sensitive components is produced by thermoelectric batteries. Are given the two-dimensional temperature fields of the electronic board during its uniform and non-uniform cooling, is carried out their comparison. As follows from the calculations results, when using a uniform overall cooling of electronic unit there is a waste of energy for the cooling 0f electronic board parts which temperature is within acceptable temperature range without the cooling system. This approach leads to the increase in the cooling capacity of used thermoelectric batteries in comparison with the desired values. This largely reduces the efficiency of heat elimination system. The use for electronic boards cooling of non-uniform local heat elimination removes this disadvantage. The obtained dependences show that in this case, the energy required to create a given temperature is smaller than when using a common uniform cooling. In this approach the temperature field of the electronic board is more uniform and the cooling is more efficient. 

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

    International Nuclear Information System (INIS)

    Tonny Lanyau; Mazleha Maskin; Mohd Fazli Zakaria; Mohmammad Suhaimi Kassim; Ahmad Nabil Abdul Rahim; Phongsakorn Prak Tom; Mohd Fairus Abdul Farid; Mohd Huzair Hussain

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-29

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

  19. Biofouling problems in freshwater cooling systems

    International Nuclear Information System (INIS)

    Rao, T.S.

    2007-01-01

    In aqueous environments, microorganisms (bacteria, algae, fungi etc.,) are attracted towards surfaces, which they readily colonise resulting in the formation of biofilms. The implications of biofouling are energy losses due to increased fluid frictional resistance and increased heat transfer resistance. The temperatures prevalent inside the condenser system provide a favorable environment for the rapid growth of microorganisms. This results in thick slime deposit, which is responsible for heat transfer losses, thereby enhancing aggregation of deposits on the material surface and induces localised corrosion. There have been instances of increased capital costs due to premature replacement of equipment caused by severe under deposit corrosion due to biofouling. Moreover, fouling of service water systems of nuclear power plants is of concern, because it reduces the heat transfer capacity during an emergency or an accident. The growth of microbial films (slimes) a few tens of microns thick, in a condenser tube is sufficient to induce microbiologically influenced corrosion and cause irreparable damage to the condenser tubes and other structural materials. The down time costs to power plant due to condenser fouling and corrosion are quite large. This paper presents the author's experience in biofouling and corrosion problems in various power plants cooled by freshwater. (author)

  20. A safety design approach for sodium cooled fast reactor core toward commercialization in Japan

    International Nuclear Information System (INIS)

    Kubo, Shigenobu

    2012-01-01

    JAEA’s safety approach for SFR core design is based on defence‐in‐depth concept, which includes DBAs and DECs (prevention and mitigation): • The reactor core is designed to have inherent reactivity feedback characteristics with negative power coefficient. • Operation temperature range is set sufficiently below the coolant boiling temperature so as to avoid coolant boiling against anticipated operational occurrences and DBAs. • If the plant state deviates from operational states, the safe reactor shutdown is achieved by automatic insertion of control rods. 2 active reactor shutdown systems are provided. • Failure of active reactor shutdown is assumed in a design extension condition . Passive shutdown capability is provided by SASS under such condition. • As a design extension condition, core disruptive accident is assumed. In order to prevent severe mechanical energy release which might cause containment function failure, core sodium void worth is limited below 6 dollars and molten fuel discharge capability is utilized by FAIDUS. (author)

  1. Restraint system for core elements of a reactor core

    International Nuclear Information System (INIS)

    Class, G.

    1975-01-01

    In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

  2. Shivering heat production and core cooling during head-in and head-out immersion in 17 degrees C water.

    Science.gov (United States)

    Pretorius, Thea; Cahill, Farrell; Kocay, Sheila; Giesbrecht, Gordon G

    2008-05-01

    Many cold-water scenarios cause the head to be partially or fully immersed (e.g., ship wreck survival, scuba diving, cold-water adventure swim racing, cold-water drowning, etc.). However, the specific effects of head cold exposure are minimally understood. This study isolated the effect of whole-head submersion in cold water on surface heat loss and body core cooling when the protective shivering mechanism was intact. Eight healthy men were studied in 17 degrees C water under four conditions: the body was either insulated or exposed, with the head either out of the water or completely submersed under the water within each insulated/exposed subcondition. Submersion of the head (7% of the body surface area) in the body-exposed condition increased total heat loss by 11% (P < 0.05). After 45 min, head-submersion increased core cooling by 343% in the body-insulated subcondition (head-out: 0.13 +/- 0.2 degree C, head-in: 0.47 +/- 0.3 degree C; P < 0.05) and by 56% in the body-exposed subcondition (head-out: 0.40 +/- 0.3 degree C and head-in: 0.73 +/- 0.6 degree C; P < 0.05). In both body-exposed and body-insulated subconditions, head submersion increased the rate of core cooling disproportionally more than the relative increase in total heat loss. This exaggerated core-cooling effect is consistent with a head cooling induced reduction of the thermal core, which could be stimulated by cooling of thermosensitive and/or trigeminal receptors in the scalp, neck, and face. These cooling effects of head submersion are not prevented by shivering heat production.

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

    International Nuclear Information System (INIS)

    El Ganaoui, K.

    2006-09-01

    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)

  4. Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

    International Nuclear Information System (INIS)

    Ginsberg, T.; Chen, J.C.

    1984-01-01

    Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench (removal of stored energy from initial temperature to saturation temperature) of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

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

    International Nuclear Information System (INIS)

    Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2003-02-01

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

  6. Dynamic Heat Storage and Cooling Capacity of a Concrete Deck with PCM and Thermally Activated Building System

    DEFF Research Database (Denmark)

    Pomianowski, Michal Zbigniew; Heiselberg, Per; Jensen, Rasmus Lund

    2012-01-01

    This paper presents a heat storage and cooling concept that utilizes a phase change material (PCM) and a thermally activated building system (TABS) implemented in a hollow core concrete deck. Numerical calculations of the dynamic heat storage capacity of the hollow core concrete deck element...... in the article highlight the potential of using TABS and PCM in a prefabricated concrete deck element....

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  8. Safety aspects of LMR [liquid metal-cooled reactor] core design

    International Nuclear Information System (INIS)

    Cahalan, J.E.

    1986-01-01

    Features contributing to increased safety margins in liquid metal-cooled reactor (LMR) design are identified. The technical basis is presented for the performance of a pool-type reactor system with an advanced metallic alloy fuel in unprotected accidents. Results are presented from analyses of anticipated transients without scram, including loss-of-flow (LOF), transient overpower (TOP), and loss-of-heat-sink (LOHS) accidents

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  10. Computational Fluid Dynamics Analysis of an Evaporative Cooling System

    Directory of Open Access Journals (Sweden)

    Kapilan N.

    2016-11-01

    Full Text Available The use of chlorofluorocarbon based refrigerants in the air-conditioning system increases the global warming and causes the climate change. The climate change is expected to present a number of challenges for the built environment and an evaporative cooling system is one of the simplest and environmentally friendly cooling system. The evaporative cooling system is most widely used in summer and in rural and urban areas of India for human comfort. In evaporative cooling system, the addition of water into air reduces the temperature of the air as the energy needed to evaporate the water is taken from the air. Computational fluid dynamics is a numerical analysis and was used to analyse the evaporative cooling system. The CFD results are matches with the experimental results.

  11. Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

    International Nuclear Information System (INIS)

    You, WuSeung; Hong, Ser Gi

    2014-01-01

    The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

  12. Modeling of Nonlinear Marine Cooling Systems with Closed Circuit Flow

    DEFF Research Database (Denmark)

    Hansen, Michael; Stoustrup, Jakob; Bendtsen, Jan Dimon

    2011-01-01

    We consider the problem of constructing a mathematical model for a specific type of marine cooling system. The system in question is used for cooling the main engine and main engine auxiliary components, such as diesel generators, turbo chargers and main engine air coolers for certain classes...

  13. Heat pump system with selective space cooling

    Science.gov (United States)

    Pendergrass, J.C.

    1997-05-13

    A reversible heat pump provides multiple heating and cooling modes and includes a compressor, an evaporator and heat exchanger all interconnected and charged with refrigerant fluid. The heat exchanger includes tanks connected in series to the water supply and a condenser feed line with heat transfer sections connected in counterflow relationship. The heat pump has an accumulator and suction line for the refrigerant fluid upstream of the compressor. Sub-cool transfer tubes associated with the accumulator/suction line reclaim a portion of the heat from the heat exchanger. A reversing valve switches between heating/cooling modes. A first bypass is operative to direct the refrigerant fluid around the sub-cool transfer tubes in the space cooling only mode and during which an expansion valve is utilized upstream of the evaporator/indoor coil. A second bypass is provided around the expansion valve. A programmable microprocessor activates the first bypass in the cooling only mode and deactivates the second bypass, and vice-versa in the multiple heating modes for said heat exchanger. In the heating modes, the evaporator may include an auxiliary outdoor coil for direct supplemental heat dissipation into ambient air. In the multiple heating modes, the condensed refrigerant fluid is regulated by a flow control valve. 4 figs.

  14. The stochastic-cooling system for COSY-Juelich

    International Nuclear Information System (INIS)

    Brittner, P.; Danzglock, R.; Hacker, H.U.; Maier, R.; Pfister, U.; Prasuhn, D.; Singer, H.; Spiess, W.; Stockhorst, H.

    1991-01-01

    The cooling in the Cooler Synchrotron COSY will work in the ranges: Band 1: 1 to 1.8 GHz, Band 2: 1.8 to 3 GHz. The system allows cooling in the energy range of 0.8 to 2.5 GeV. The stochastic-cooling system is under development. Cooling characteristics have been calculated. The tanks are similar to those of the CERN-AC. But the COSY parameters have required changes of the tank design. Active RF components have been developed for COSY. Measured results are presented

  15. Improved core protection calculator system algorithm

    International Nuclear Information System (INIS)

    Yoon, Tae Young; Park, Young Ho; In, Wang Kee; Bae, Jong Sik; Baeg, Seung Yeob

    2009-01-01

    Core Protection Calculator System (CPCS) is a digitized core protection system which provides core protection functions based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels which adapted a two out of four trip logic. CPCS algorithm improvement for the newly designed core protection calculator system, RCOPS (Reactor COre Protection System), is described in this paper. New features include the improvement of DNBR algorithm for thermal margin, the addition of pre trip alarm generation for auxiliary trip function, VOPT (Variable Over Power Trip) prevention during RPCS (Reactor Power Cutback System) actuation and the improvement of CEA (Control Element Assembly) signal checking algorithm. To verify the improved CPCS algorithm, CPCS algorithm verification tests, 'Module Test' and 'Unit Test', would be performed on RCOPS single channel facility. It is expected that the improved CPCS algorithm will increase DNBR margin and enhance the plant availability by reducing unnecessary reactor trips

  16. Integrated systems for power plant cooling and wastewater management

    International Nuclear Information System (INIS)

    Haith, D.A.

    1975-01-01

    The concept of integrated management of energy and water resources, demonstrated in hydropower development, may be applicable to steam-generated power, also. For steam plants water is a means of disposing of a waste product, which is unutilized energy in the form of heat. One framework for the evolution of integrated systems is the consideration of possible technical linkages between power plant cooling and municipal wastewater management. Such linkages include the use of waste heat as a mechanism for enhancing wastewater treatment, the use of treated wastewater as make-up for evaporative cooling structures, and the use of a pond or reservoir for both cooling and waste stabilization. This chapter reports the results of a systematic evaluation of possible integrated systems for power plant cooling and waste water management. Alternatives were analyzed for each of three components of the system--power plant cooling (condenser heat rejection), thermally enhanced waste water treatment, and waste water disposal. Four cooling options considered were evaporative tower, open cycle, spray pond, and cooling pond. Three treatment alternatives considered were barometric condenser-activated sludge, sectionalized condenser-activated sludge, and cooling/stabilization pond. Three disposal alternatives considered were ocean discharge, land application (spray irrigation), and make-up (for evaporative cooling). To facilitate system comparisons, an 1100-MW nuclear power plant was selected. 31 references

  17. Experiments on novel solar heating and cooling system

    International Nuclear Information System (INIS)

    Wang Yiping; Cui Yong; Zhu Li; Han Lijun

    2008-01-01

    Solar heating and nocturnal radiant cooling techniques are united to produce a novel solar heating and cooling system. The radiant panel with both heating and cooling functions can be used as structural materials for the building envelope, which realizes true building integrated utilization of solar energy. Based on the natural circulation principle, the operation status can be changed automatically between the heating cycle and the cooling cycle. System performances under different climate conditions using different covers on the radiant panel are studied. The results show that the novel solar heating and cooling system has good performance of heating and cooling. For the no cover system, the daily average heat collecting efficiency is 52% with the maximum efficiency of 73%, while at night, the cooling capacity is about 47 W/m 2 on a sunny day. On a cloudy day, the daily average heat collecting efficiency is 47% with the maximum of 84%, while the cooling capacity is about 33 W/m 2 . As a polycarbonate (PC) panel or polyethylene film are used as covers, the maximum heat collecting efficiencies are 75% and 72% and the daily average heat collecting efficiencies are 61% and 58%, while the cooling capacities are 50 W/m 2 and 36 W/m 2 , respectively

  18. Application of fuzzy control in cooling systems save energy design

    Energy Technology Data Exchange (ETDEWEB)

    Chen, M.L.; Liang, H.Y. [Chienkuo Technology Univ., Changhua, Taiwan (China). Dept. of Electrical Engineering

    2005-07-01

    A fuzzy logic programmable logic controller (PLC) was used to control the cooling systems of frigorific equipment. Frigorific equipment is used to move unwanted heat outside of building in order to control indoor temperatures. The aim of the fuzzy logic PLC was to improve the energy efficiency of the cooling system. Control of the cooling pump and cooling tower in the system was based on the water temperature of the condenser during frigorific system operation. A human computer design for the cooling system control was used to set speeds and to automate and adjust the motor according to the fuzzy logic controller. It was concluded that if fuzzy logic controllers are used with all components of frigorific equipment, energy efficiency will be significantly increased. 5 refs., 3 tabs., 9 figs.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  20. VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE

    Directory of Open Access Journals (Sweden)

    NAM-IL TAK

    2013-11-01

    Full Text Available For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR, intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI and the AGREE code of the University of Michigan (U of M. One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

  1. Core Flight System (CFS) Integrated Development Environment

    Data.gov (United States)

    National Aeronautics and Space Administration — The purpose of this project is to create an Integrated Development Environment (IDE) for the Core Flight System (CFS) software to reduce the time it takes to...

  2. Core Flight System Satellite Starter Kit

    Data.gov (United States)

    National Aeronautics and Space Administration — The Core Flight System Satellite Starter Kit (cFS Kit) will allow a small satellite or CubeSat developer to rapidly develop, deploy, test, and operate flight...

  3. Application of a bistable convection loop to LMFBR [liquid metal fast breeder reactor] emergency core cooling

    International Nuclear Information System (INIS)

    Anand, G.; Christensen, R.N.

    1990-01-01

    The concept of passive safety features for nuclear reactors has been developed in recent years and has gained wide acceptance. A literature survey of current reactors with passive features indicates that these reactors have some passive features but still do not fully meet the design objectives. Consider a current liquid-metal reactor design like PRISM. During normal operation, liquid sodium enters the reactor at ∼395 degree C and exits at ∼550 degree C. In the event of loss of secondary cooling with or without scram, the primary coolant (liquid sodium) initially acts as a heat sink and its temperature increases. For events without scram, the negative reactivity induced by the increase in temperature shuts the reactor down. When the average temperature of the sodium reaches ∼600 to 650 degree C, it overflows from the reactor vessel, activating the auxiliary cooling system. The auxiliary cooling system uses natural circulation of air around the reactor guard vessel. An alternative to the current design incorporates a bistable convection loop (BCL). The incorporation of the BCL concept remarkably improves the safety of the nuclear reactors. Application of the BCL concept to liquid-metal fast breeder reactors is described in this paper

  4. Validation of reactor core protection system

    International Nuclear Information System (INIS)

    Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

    2008-01-01

    Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

  5. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  6. Multi-core fiber undersea transmission systems

    DEFF Research Database (Denmark)

    Nooruzzaman, Md; Morioka, Toshio

    2017-01-01

    Various potential architectures of branching units for multi-core fiber undersea transmission systems are presented. It is also investigated how different architectures of branching unit influence the number of fibers and those of inline components.......Various potential architectures of branching units for multi-core fiber undersea transmission systems are presented. It is also investigated how different architectures of branching unit influence the number of fibers and those of inline components....

  7. Fundamental research on the cooling characteristic of passive containment cooling system

    International Nuclear Information System (INIS)

    Kawakubo, M.; Kikura, H.; Aritomi, M.; Inaba, N.; Yamauchi, T.

    2004-01-01

    The objective of this experimental study is to clarify the heat transfer characteristics of the Passive Containment Cooling System (PCCS) with vertical heat transfer tubes for investigating the influence of non-condensable gas on condensation. Furthermore, hence we obtained new experimental correlation formula to calculate the transients in system temperature and pressure using the simulation program of the PCCS. The research was carried out using a forced circulation experimental loop, which simulates atmosphere inside PCCS with vertical heat transfer tubes if a loss of coolant accident (LOCA) occurs. The experimental facility consists of cooling water supply systems, an orifice flowmeter, and a tank equipped with the heat transfer pipe inside. Cooling water at a constant temperature is injected to the test part of heat transfer pipe vertically installed in the tank by forced circulation. At that time, the temperature of the cooling water between inlet and outlet of the pipe was measured to calculate the overall heat transfer coefficient between the cooling water and atmosphere in the tank. Thus, the heat transfer coefficient between heat transfer surface and the atmosphere in the tank considering the influence of the non-condensable gas was clarified. An important finding of this study is that the amount of condensation in the steamy atmosphere including non-condensable gas depends on the cooling water Reynolds number, especially the concentration of non-condensable gas that has great influence on the amount of condensation. (authors)

  8. Percutaneous radiofrequency ablation of osteoid osteoma using cool-tip electrodes without the cooling system

    International Nuclear Information System (INIS)

    Miyazaki, Masaya; Miyazaki, Akiko; Nakajima, Takahito; Koyama, Yoshinori; Shinozaki, Tetsuya; Endo, Keigo; Aoki, Jun

    2011-01-01

    The aim of this study was to evaluate the efficacy of percutaneous radiofrequency ablation (RFA) for osteoid osteoma (OO) using cool-tip electrodes without the cooling system. A total of 17 patients (13 males, 4 females; mean age 19.1 years; range 7-49 years) with OO (tibia, n=7; femur, n=5; acetabulum, n=2; radius, n=1; talus, n=1; lumbar spine, n=1) underwent RFA. Using a cool-tip electrode without the cooling system, the lesion was heated to 90degC for 4 or 5 min. Procedures were considered technically successful if the electrode was placed into the nidus and the target temperature was reached and maintained for at least 4 min. Clinical success of the treatment was defined as complete or partial pain relief after RFA. All procedures were considered technically successful, although two patients encountered complications (pes equinus contracture, skin burn). Altogether, 16 of the 17 patients (94.1%) achieved complete or partial pain relief after primary RFA. Two patients had pain recurrence, with one of them treated successfully with a second RFA. The overall clinical success rate was 88.2%. Histological findings confirmed the presence of OO in 13 patients (76.5%). Percutaneous RFA of OO using cool-tip electrodes without the cooling system is a safe, effective procedure. (author)

  9. PERFORMANCE EVALUATION OF CEILING RADIANT COOLING SYSTEM IN COMPOSITE CLIMATE

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Anuj [Malaviya National Institute of Technology (MNIT), Jaipur, India; Mathur, Jyotirmay [Malaviya National Institute of Technology (MNIT), Jaipur, India; Bhandari, Mahabir S [ORNL

    2015-01-01

    Radiant cooling systems are proving to be an energy efficient solution due to higher thermal capacity of cooling fluid especially for the buildings that require individual zone controls and where the latent loads are moderate. The Conventional air conditioners work at very low temperature i.e.5-8 c (refrigerant evaporator inlet) while the radiant cooling systems, also referred as high temperature cooling system, work at high temperatures i.e. 14-18 c. The radiant cooling systems can maintain lower MRT (Mean Radiant Temperature) as ceiling panels maintain uniform temperature gradient inside room and provide higher human comfort. The radiant cooling systems are relatively new systems and their operation and energy savings potential are not quantified for a large number of buildings and operational parameters. Moreover, there are only limited numbers of whole building simulation studies have been carried out for these systems to have a full confidence in the capability of modelling tools to simulate these systems and predict the impact of various operating parameters. Theoretically, savings achieve due to higher temperature set point of chilled water, which reduces chiller-running time. However, conventional air conditioner runs continuously to maintain requisite temperature. In this paper, experimental study for performance evaluation of radiant cooling system carried out on system installed at Malaviya National Institute of Technology Jaipur. This paper quantifies the energy savings opportunities and effective temperature by radiant cooling system at different chilled water flow rates and temperature range. The data collected/ analysed through experimental study will used for calibration and validation of system model of building prepared in building performance simulation software. This validated model used for exploring optimized combinations of key parameters for composite climate. These optimized combinations will used in formulation of radiant cooling system

  10. Development of CANDU core monitoring system

    International Nuclear Information System (INIS)

    Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H.

    2003-01-01

    The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1

  11. Development of CANDU core monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H. [Institute for Advanced Engineering, Yongin (Korea, Republic of)

    2003-07-01

    The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1.

  12. Web-based Core Design System Development

    International Nuclear Information System (INIS)

    Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

    2011-01-01

    The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

  13. Improvement of Cooling Performance of a Compact Thermoelectric Air Conditioner Using a Direct Evaporative Cooling System

    Science.gov (United States)

    Tipsaenporm, W.; Lertsatitthanakorn, C.; Bubphachot, B.; Rungsiyopas, M.; Soponronnarit, S.

    2012-06-01

    This paper presents the results of tests carried out to investigate the potential application of a direct evaporative cooling (DEC) system for improving the performance of a compact thermoelectric (TE) air conditioner. The compact TE air conditioner is composed of three TE modules. The cold and hot sides of the TE modules were fixed to rectangular fin heat sinks. The DEC system produced cooling air that was used to assist the release of heat from the heat sinks at the hot side of the TE modules. The results showed that the cooling air dry bulb temperature from the DEC system achieved drops of about 5.9°C in parallel with about a 33.4% rise in relative humidity. The cooling efficiency of the DEC system varies between 72.1% and 81.5%. It increases the cooling capacity of the compact TE air conditioner from 53.0 W to 74.5 W. The 21.5 W (40.6%) increase represents the difference between the compact air conditioner operating with ambient air flowing through the TE module's heat sinks, and the compact air conditioner operating with the cooler air from the DEC system flowing through the TE module's heat sinks. In both scenarios, electric current of 4.5 A was supplied to the TE modules. It also has been experimentally proven that the coefficient of performance (COP) of the compact TE air conditioner can be improved by up to 20.9% by incorporating the DEC system.

  14. 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)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    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

  15. Calculation of the neutron noise induced by periodic deformations of a large sodium-cooled fast reactor core

    International Nuclear Information System (INIS)

    Zylbersztejn, F.; Tran, H.N.; Pazsit, I.; Filliatre, P.; Jammes, C.

    2014-01-01

    The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed. (authors)

  16. Cooling systems for waste heat. Cooling systems, review and selection criteria. Kuehlsysteme fuer Abwaerme. Kuehlsysteme, Ueberblick und Auswahlkriterien

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, W. (Jaeggi, Wallisellen (Switzerland))

    1990-05-01

    In many areas of ventilation, air-conditioning and refrigeration engineering, chemical and process engineering and energy production waste heat occurs. If a reduction in energy losses or heat recovery is not possible waste heat has to be drawn off through cooling systems. For this the following systems can be used: dry cooling systems, dry cooler with spray system, open-cycle wet cooler, hybrid dry cooler, and closed-cycle wet cooler. Particularly hybrid cooling systems can give acceptable solutions when the results with other systems are only unsatisfactory. (BWI).

  17. Experimental and numerical CHT-investigations of cooling structures formed by lost cores in cast housings for optimal heat transfer

    Science.gov (United States)

    Kohlstädt, S.; Vynnycky, M.; Gebauer-Teichmann, A.

    2018-05-01

    This paper investigates the cooling performance of six different lost core designs for automotive cast houses with regard to their cooling efficiency. For this purpose, the conjugate heat transfer (CHT) solver, chtMultiregion, of the freely available CFD-toolbox OpenFOAM in its implementation of version 2.3.1 is used. The turbulence contribution to the Navier-Stokes equations is accounted for by using the RANS Menter SST k - ω model. The results are validated for one of the geometries by comparing with experimental data. Of the six investigated cooling structures, the one that forces the fluid flow to change its direction the most produces the lowest temperatures on the surface of the cast housing. This good cooling performance comes at the price of the highest pressure loss in the cooling fluid and hence increased pump power. It is also found that the relationship between performance and pressure drop is by no means generally linear. Slight changes in the design can lead to a structure which cools almost as well, but at much decreased pressure loss. Regarding the absolute values, the simulations showed that the designed cooling structures are suitable for handling the cooling requirements in the particular applications and that the maximum temperature stays below the critical limits of the electronic components.

  18. Numerical study of a novel dew point evaporative cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Riangvilaikul, B.; Kumar, S. [Energy Field of Study, School of Environment, Resources and Development, Asian Institute of Technology, P.O. Box 4, Klong Luang, Pathumthani 12120 (Thailand)

    2010-11-15

    Dew point evaporative cooling system is an alternative to vapor compression air conditioning system for sensible cooling of ventilation air. This paper presents the theoretical performance of a novel dew point evaporative cooling system operating under various inlet air conditions (covering dry, moderate and humid climate) and influence of major operating parameters (namely, velocity, system dimension and the ratio of working air to intake air). A model of the dew point evaporative cooling system has been developed to simulate the heat and mass transfer processes. The outlet air conditions and system effectiveness predicted by the model using numerical method for known inlet parameters have been validated with experimental findings and with recent literature. The model was used to optimize the system parameters and to investigate the system effectiveness operating under various inlet air conditions. (author)

  19. Closed-cycle cooling systems for nuclear power plants

    International Nuclear Information System (INIS)

    Santini, Lorenzo

    2006-01-01

    The long experience in the field of closed-cycle cooling systems and high technological level of turbo machines and heat exchangers concurs to believe in the industrial realizability of nuclear systems of high thermodynamic efficiency and intrinsic safety [it

  20. Solar heating and cooling technical data and systems analysis

    Science.gov (United States)

    Christensen, D. L.

    1977-01-01

    The research activities described herein were concentrated on the areas of economics, heating and cooling systems, architectural design, materials characteristics, climatic conditions, educational information packages, and evaluation of solar energy systems and components.

  1. RAMI analysis for DEMO HCPB blanket concept cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

    2015-10-15

    Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

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

    International Nuclear Information System (INIS)

    Menke, H.

    1988-01-01

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

  3. Materials for innovative lead alloy cooled nuclear systems: Overview

    International Nuclear Information System (INIS)

    Mueller, Georg; Weisenburger, Alfons; Fetzer, Renate; Heinzel, Annette; Jianu, Adrian

    2015-01-01

    One of the most challenging issues for all future innovative nuclear systems including Gen IV reactors are materials. The selection of the structural materials determines the design which has to consider the properties and the availability of the materials. Beside general requirements for material properties that are common for all fast reactor types specific issues arise from coolant compatibility. The high solubility of steel alloying elements in liquid Pb-alloys at reactor relevant temperatures is clearly detrimental. Therefore, all steels that are considered as structural materials have to be protected by dissolution barriers. The most common barriers for steels under consideration are oxide scales that form in situ during operation. However, increasing the temperature above 500 deg. C will result either in dissolution attack or in enhanced oxidation. For higher temperatures additional barriers like alumina forming surface alloys are discussed and investigated. Mechanical loads like creep stress and fretting will act on the steels. These mechanical loads will interact with the coolant and can increase the negative effects. For a LFR (Lead Fast Reactor) Demonstrator and MYHRRA (ADS) austenitic steels (316L) are selected for most in core components. The 15-15Ti is the choice for the fuel cladding of MYHRRA and a Pb cooled demonstrator. For an industrial LFR (Lead Fast Reactor) the ferritic martensitic steel T91 was selected as fuel clad material due to its improved irradiation resistance. T91 is in both designs the material to be used for the heat exchanger. Surface alloying with alumina forming alloys is considered to assure material functionality at higher temperatures and is therefore selected for fuel cladding of the ELFR and the heat exchanger tubes. This presentation will give an overview on the selected materials for innovative Pb alloy cooled nuclear systems considering, beside pure compatibility, the influence of mechanical interaction like creep and

  4. Thermal Hydraulic Analysis of RPV Support Cooling System for HTGR

    International Nuclear Information System (INIS)

    Min Qi; Wu Xinxin; Li Xiaowei; Zhang Li; He Shuyan

    2014-01-01

    Passive safety is now of great interest for future generation reactors because of its reduction of human interaction and avoidance of failures of active components. reactor pressure vessel (RPV) support cooling system (SCS) for high temperature gas-cooled reactor (HTGR) is a passive safety system and is used to cool the concrete seats for the four RPV supports at its bottom. The SCS should have enough cooling capacity to ensure the temperature of the concrete seats for the supports not exceeding the limit temperature. The SCS system is composed of a natural circulation water loop and an air cooling tower. In the water loop, there is a heat exchanger embedded in the concrete seat, heat is transferred by thermal conduction and convection to the cooling water. Then the water is cooled by the air cooler mounted in the air cooling tower. The driving forces for water and air are offered by the density differences caused by the temperature differences. In this paper, the thermal hydraulic analysis for this system was presented. Methods for decoupling the natural circulation and heat transfer between the water loop and air flow were introduced. The operating parameters for different working conditions and environment temperatures were calculated. (author)

  5. Evaluation of two cooling systems under a firefighter coverall

    NARCIS (Netherlands)

    Teunissen, L.P.J.; Wang, L.C.; Chou, S.N.; Huang, C.; Jou, G.T.; Daanen, H.A.M.

    2014-01-01

    Firemen often suffer from heat strain. This study investigated two chest cooling systems for use under a firefighting suit. In nine male subjects, a vest with water soaked cooling pads and a vest with water perfused tubes were compared to a control condition. Subjects performed 30 min walking and 10

  6. Study on the seismic verification test program on the experimental multi-purpose high-temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Taketani, K.; Aochi, T.; Yasuno, T.; Ikushima, T.; Shiraki, K.; Honma, T.; Kawamura, N.

    1978-01-01

    The paper describes a program of experimental research necessary for qualitative and quantitative determination of vibration characteristics and aseismic safety on structure of reactor core in the multipurpose high temperature gas-cooled experimental reactor (VHTR Experimental Reactor) by the Japan Atomic Energy Research Institute

  7. Conceptual design study on simplified and safer cooling systems for sodium cooled FBRs

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Shimakawa, Yoshio; Ishikawa, Hiroyasu; Kubota, Kenichi; Kobayashi, Jun; Kasai, Shigeo

    2000-06-01

    The objective of this study is to create the FBR plant concepts increasing economy and safety for the Phase-I 'Feasibility Studies on Commercialized Fast Reactor System'. In this study, various concepts of simplified 2ry cooling system for sodium cooled FBRs are considered and evaluated from the view points of technological feasibility, economy, and safety. The concepts in the study are considered on the basis of the following points of view. 1. To simplify 2ry cooling system by moderating and localizing the sodium-water reaction in the steam generator of the FBRs. 2. To simplify 2ry cooling system by eliminating the sodium-water reaction using integrated IHX-SG unit. 3. To simplify 2ry cooling system by eliminating the sodium-water reaction using a power generating system other than the steam generator. As the result of the study, 12 concepts and 3 innovative concepts are proposed. The evaluation study for those concepts shows the following technical prospects. 1. 2 concepts of integrated IHX-SG unit can eliminate the sodium-water reaction. Separated IHX and SG tubes unit using Lead-Bismuth as the heat transfer medium. Integrated IHX-SG unit using copper as the heat transfer medium. 2. Cost reduction effect by simplified 2ry cooling system using integrated IHX-SG unit is estimated 0 to 5%. 3. All of the integrated IHX-SG unit concepts have more weight and larger size than conventional steam generator unit. The weight of the unit during transporting and lifting would limit capacity of heat transfer system. These evaluation results will be compared with the results in JFY 2000 and used for the Phase-II study. (author)

  8. Effect of in-core instrumentation mounting location on external reactor vessel cooling

    International Nuclear Information System (INIS)

    Suh, Jungsoo; Ha, Huiun

    2017-01-01

    Highlights: • Numerical simulations were conducted for the evaluation of an IVR-ERVC application. • The ULPU-V experiment was simulated for the validation of numerical method. • The effect of ICI mounting location on an IVR-ERVC application was investigated. • TM-ICI is founded to be superior to BM-ICI for successful application of IVR-ERVC. - Abstract: The effect of in-core instrumentation (ICI) mounting location on the application of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC), used to mitigate severe accidents in which the nuclear fuel inside the reactor vessel becomes molten, was investigated. Numerical simulations of the subcooled boiling flow within an advanced pressurized-water reactor (PWR) in IVR-ERVC applications were conducted for the cases of top-mounted ICI (TM-ICI) and bottom-mounted ICI (BM-ICI), using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. To validate the numerical method for IVR applications, numerical simulations of ULPU-V experiments were also conducted. The BM-ICI reactor vessel was modeled using a simplified design of an advanced PWR with BM-ICI; the TM-ICI counterpart was modeled by removing the ICI parts from the original geometry. It was found that TM-ICI was superior to BM-ICI for successful application of IVR-ERVC. For the BM-ICI case, the flow field was complicated because of the existence of ICIs and a significant temperature gradient was observed near the ICI nozzles on the lower part of the reactor vessel, where the ICIs were attached. These observations suggest that the existence of ICI below the reactor vessel hinders reactor vessel cooling.

  9. Functional requirements for core surveillance systems

    International Nuclear Information System (INIS)

    Andersson, T.

    2000-01-01

    Operating experience at Ringhals-2 has demonstrated the feasibility of a mixed core surveillance system comprised of fixed in-core detectors combined with the original movable detector system. A small number of fixed in-core detectors provide continuous measurement of the thermal margins while the movable detectors are used mainly at start-up to verify the expected power distribution. Reactor noise diagnostics and neural networks can further improve the monitoring system. The reliability of the movable detector system can be improved by mechanical simplification. Wear and maintenance costs are lowered if the required flux-mapping frequency is reduced. Improved computer codes make the measurement uncertainties less dependent on the number of instrumented positions. A mixed system requires new types of technical specifications. (author)

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

    International Nuclear Information System (INIS)

    Pham Nhu Viet Ha; Min Jae Lee; Sunghwan Yun; Sang Ji Kim

    2015-01-01

    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)

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

    International Nuclear Information System (INIS)

    Hahn, Oliver; Martizzi, Davide; Wu, Hao-Yi

    2017-01-01

    We present the rhapsody-g suite of cosmological hydrodynamic zoom simulations of 10 massive galaxy clusters at the M vir ~10 15 M ⊙ scale. These simulations include cooling and subresolution 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 active galactic nuclei feedback. For cluster scaling relations, we find that the simulations match well the M 500 –Y 500 scaling of Planck Sunyaev–Zeldovich clusters but deviate somewhat from the observed X-ray luminosity and temperature scaling relations in the sense of being slightly too bright and too cool at fixed mass, respectively. Stars are produced at an efficiency consistent with abundance-matching constraints and central galaxies have star formation rates consistent with recent observations. In conclusion, while our simulations thus match various key properties remarkably well, we conclude that the shortcomings strongly suggest an important role for non-thermal processes (through feedback or otherwise) or thermal conduction in shaping the intracluster medium.

  12. Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

    OpenAIRE

    Verma, Vasudha

    2017-01-01

    Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

  13. Analysis of a solid desiccant cooling system with indirect evaporative cooling

    DEFF Research Database (Denmark)

    Bellemo, Lorenzo

    investigates the performance of a solid desiccant cooling system implementing in-direct evaporative cooling processes. The aim is to quantify the system thermal and electrical performance for varying component dimensions and operating conditions, and to identify its range of applicability. This information...... evaporative cooler. Detailed steady state numerical models are developed and implemented in MATLAB. The models need to be accurate and require low computational effort, for analysing the internal heat and mass transfer processes, as well as carrying out repetitive design and optimization simulations......-to-air heat exchanger for enhancing cooling capacity and thermal performance. The system perfor-mance is investigated considering regeneration temperatures between 50 ºC and 90 ºC, which enable low temperature heat sources, such as solar energy or waste heat, to be used. The effects of several geometrical...

  14. Heat Driven Cooling in District Energy Systems; Vaermedriven Kyla

    Energy Technology Data Exchange (ETDEWEB)

    Rydstrand, Magnus; Martin, Viktoria; Westermark, Mats [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Chemical Engineering and Technology

    2004-07-01

    high costs. However heat sinks are unavoidable from a system perspective and there are potential cost savings since a low-pressure steam turbines will not be required if heat driven cooling is implemented. The fuel utilization for some technologies (not necessarily the best technology) was evaluated in two different scenarios: 1) with electricity production from coal; and 2) with electricity production from natural gas. It is shown in the scenarios that the heat driven cooling technologies give lower fuel consumption as compared producing electricity as an intermediate product before cooling is produced. Further it should be noted that electricity is produced, not consumed, if heat is used directly for the production of cooling. We claim that cost effective solutions for district heat driven chillers and/or combined production of electricity and district cooling can be found in all climates with high enough density of heating and cooling demands. It was found that district heat driven chillers can be very energy efficient in warm and humid climates since desiccant systems are an effective way of handling latent cooling loads. In dry climates, with low latent loads, water distributed cooling has a large potential and absorption cooling will give high fuel utilization seen from a system perspective. In climates where water shortage is a problem it is possible that the temperature lift of the conventional absorption chiller has to be increased in order to be able to use dry cooling towers. The temperature lift can be increased by changing the chiller design or by using a different working pair. Heat driven cooling can be integrated into an energy system in different ways. In USA and Japan, district heating is not well developed. Instead small, distributed combined heat and power (CHP) plants with high exhaust temperatures are widespread. Cooling is often produced, in these regions, through absorption cooling (using heat from CHP) or compression chillers depending on

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

    International Nuclear Information System (INIS)

    Bi, Yuehong; Liu, Xiao; Jiang, Minghe

    2014-01-01

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

  16. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations

  17. First evidence of diffuse ultra-steep-spectrum radio emission surrounding the cool core of a cluster

    Science.gov (United States)

    Savini, F.; Bonafede, A.; Brüggen, M.; van Weeren, R.; Brunetti, G.; Intema, H.; Botteon, A.; Shimwell, T.; Wilber, A.; Rafferty, D.; Giacintucci, S.; Cassano, R.; Cuciti, V.; de Gasperin, F.; Röttgering, H.; Hoeft, M.; White, G.

    2018-05-01

    Diffuse synchrotron radio emission from cosmic-ray electrons is observed at the center of a number of galaxy clusters. These sources can be classified either as giant radio halos, which occur in merging clusters, or as mini halos, which are found only in cool-core clusters. In this paper, we present the first discovery of a cool-core cluster with an associated mini halo that also shows ultra-steep-spectrum emission extending well beyond the core that resembles radio halo emission. The large-scale component is discovered thanks to LOFAR observations at 144 MHz. We also analyse GMRT observations at 610 MHz to characterise the spectrum of the radio emission. An X-ray analysis reveals that the cluster is slightly disturbed, and we suggest that the steep-spectrum radio emission outside the core could be produced by a minor merger that powers electron re-acceleration without disrupting the cool core. This discovery suggests that, under particular circumstances, both a mini and giant halo could co-exist in a single cluster, opening new perspectives for particle acceleration mechanisms in galaxy clusters.

  18. 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.

  19. Thermodynamic analysis of cooling systems for nuclear power stations condenser

    International Nuclear Information System (INIS)

    Beck, A.

    1985-06-01

    This work is an attempt to concentrate on the thermodynamic theory, the engineering solution and the quantities of water needed for the operation of a wet as well as a wet/dry cooling towers coupled to a nuclear turbine condenser,. About two hundred variables are needed for the design of a condenser - cooling tower system. In order to make the solution fast and handy, a computer model was developed. The amount of water evaporation from cooling towers is a function of the climate conditions prevailing around the site. To achieve an authentic analysis, the meteorological data of the northern Negev was used. The total amount of water necessary to add to the system in a year time of operation is large and is a function of both the blow-down rate and the evaporation. First estimations show that the use of a combined system, wet/dry cooling tower, is beneficial in the northern Negev area. Such a system can reduce significantly the amount of wasted fresh water. Lack of international experience is the major problem in the acceptability of wet/dry cooling towers. The technology of a wet cooling tower using sea water is also discussed where no technical or engineering limitations were found. This work is an attempt to give some handy tools for making the choice of cooling systems for nuclear power plants easier

  20. Preoperational test report, primary ventilation condenser cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Clifton, F.T.

    1997-10-29

    This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System.

  1. Preoperational test report, primary ventilation condenser cooling system

    International Nuclear Information System (INIS)

    Clifton, F.T.

    1997-01-01

    This represents the preoperational test report for the Primary Ventilation Condenser Cooling System, Project W-030. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. The system uses a closed chilled water piping loop to provide offgas effluent cooling for tanks AY101, AY102, AZ1O1, AZ102; the offgas is cooled from a nominal 100 F to 40 F. Resulting condensation removes tritiated vapor from the exhaust stack stream. The piping system includes a package outdoor air-cooled water chiller with parallel redundant circulating pumps; the condenser coil is located inside a shielded ventilation equipment cell. The tests verify correct system operation and correct indications displayed by the central Monitor and Control System

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  3. CAREM 25: Suppression pool cooling and purification system

    International Nuclear Information System (INIS)

    Carlevaris, Rodolfo; Patrignani, Alberto; Vindrola, Carlos; Palmerio, Hector D.; Quiroz, Horacio; Ramilo, Lucia B.

    2000-01-01

    The suppression pool cooling and purification system has the following main functions: purify and cool water from the suppression pool, cool and send water to the residual heat extraction system, and transfer water to the fuel element transference channel. In case of Loss of Coolant Accident (LOCA), the system sends water from the suppression pool to the spray network, thus cooling and reducing pressure in the primary containment. The system has been designed in accordance with the requirements of the following standards: ANSI/ANS 52.1; ANSI/ANS 57.2; ANSI/ANS 56.2; ANSI/ANS 59.1; ANSI/ANS 58.3; ANSI/ANS 58.9; and ANSI/ANS 56.5. The design of the system fulfils all the assigned functions. (author)

  4. CAREM-25. Suppression Pool Cooling and Purification System

    International Nuclear Information System (INIS)

    Carlevaris, Rodolfo; Palmerio, D.; Patrignani, A.; Quiroz, H.; Ramilo, L.; Vindrola, C.

    2000-01-01

    The Suppression Pool Cooling and Purification System has the following main functions: purify and cool water from the Suppression Pool, cool and send water to the Residual Heat Extraction System, and transfer water to the Fuel Element Transference Channel. In case of Loss of Coolant Accident (LOCA), the system sends water from the Suppression Pool to the spray network, thus cooling and reducing pressure in the primary containment.The system has been designed in accordance with the requirements of the following standards ANSI/ANS 52.1 [1], ANSI/ANS 57.2 [2], ANSI/ANS 56.2 [3], ANSI/ANS 59.1 [4] ANSI/ANS 58.3 [5], ANSI/ANS 58.9 [6], and ANSI/ANS 56.5 [7]. The design of the system fulfils all the assigned functions

  5. Evaluation of load case ``switch-off of the high pressure pump of the emergency core cooling system``, measures of verification and in situ-test; Einstufung des Lastfalls ``Ausfall der TH-Hochdruckeinspeisepumpe``, Massnahmen zur Verifikation bis hin zum Grossversuch

    Energy Technology Data Exchange (ETDEWEB)

    Trobitz, M.; Mattheis, A. [Kernkraftwerke Gundremmingen Betriebsgesellschaft m.b.H. (Germany); Kerkhof, K.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Gurr-Beyer, C. [Buero fuer Baudynamik, Stuttgart (Germany); Hofstoetter, P. [Technischer Ueberwachungs-Verein Rheinland e.V., Koeln (Germany)

    1998-11-01

    Within the framework of periodic safety inspection of the Gundremmingen power station (RWE-Bayernwerk - KRB II), the load collectives used for the design of safety-relevant systems and components were checked for their consistency with latest updates of the design basis. It was found that there was no analytical information or study available describing a particular process and its effects, namely switch-off of the high-pressure feedwater pump of the emergency core cooling system. The paper reports the work performed for closing the gap, including preparatory analyses, accompanying measures such as vibration measurements during plant shut-down, as well as the preparation and performance of the in-situ test. The experimental results and the comparative evaluation of calculated and experimental data are presented. (orig./CB) [Deutsch] Im Rahmen der periodischen Sicherheitsueberpruefung des Kernkraftwerkes Gundremmingen (Kernkraftwerke RWE-Bayernwerk - KRB II) wurden u.a. die Lastkollektive, die zur Auslegung sicherheitstechnisch relevanter Systeme und Komponenten herangezogen wurden, auf Aktualitaet ueberprueft. Dabei zeigte sich, dass bislang fuer eine Betriebsweise - naemlich das Abschalten der Hochdruckeinspeisepumpe des nuklearen Not- und Nachkuehlsystems (TH-HD-Pumpe) - keine analytischen Untersuchungen vorliegen. Vorbetrachtungen fuer analytische Untersuchungen, begleitende Massnahmen wie Schwingungsmessungen waehrend des Anlagenstillstandes, sowie der Versuchsaufbau und die Versuchsdurchfuehrung des Anlagenversuches werden hier dargestellt. Die Ergebnisse und der Vergleich Rechnung-Messung zum Grossversuch werden in diesem Beitrag vorgestellt. (orig.)

  6. Cooling System Design Options for a Fusion Reactor

    Science.gov (United States)

    Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

    1997-06-01

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

  7. Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

    Science.gov (United States)

    Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

    Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

  8. 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.

  9. Armor systems including coated core materials

    Science.gov (United States)

    Chu, Henry S [Idaho Falls, ID; Lillo, Thomas M [Idaho Falls, ID; McHugh, Kevin M [Idaho Falls, ID

    2012-07-31

    An armor system and method involves providing a core material and a stream of atomized coating material that comprises a liquid fraction and a solid fraction. An initial layer is deposited on the core material by positioning the core material in the stream of atomized coating material wherein the solid fraction of the stream of atomized coating material is less than the liquid fraction of the stream of atomized coating material on a weight basis. An outer layer is then deposited on the initial layer by positioning the core material in the stream of atomized coating material wherein the solid fraction of the stream of atomized coating material is greater than the liquid fraction of the stream of atomized coating material on a weight basis.

  10. Analysis Of The Heat Exchanger Capability At One Line Cooling System Operation Mode Of The RSG-GAS

    International Nuclear Information System (INIS)

    Dibyo, Sukmanto; Kuntoro, Iman

    2000-01-01

    In the frame of minimizing the operation lost of the RSG-GAS reactor, operation using one line cooling system at certain power range is being evaluated. Analysis the performance of cooling system for determining maximum power should be carried out. Analysis was carried out based on heat exchanger calculation using actual operation data. Constraints imposed to the analysis are that inlet cooling system to the reactor core shall be less than 42 o C. The result shows that by using one line of primary and secondary coolant flow of 1780 m exp. 3/hr and 2000 m 3 /hr and secondary coolant temperature from the cooling tower of 38 o C, the primary coolant to the core will be reach 42 o C if reactor operated at power of 16 MW

  11. Active cooling system for Tokamak in-vessel operation manipulator

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Jianjun, E-mail: yuanjj@sjtu.edu.cn; Chen, Tan; Li, Fashe; Zhang, Weijun; Du, Liang

    2015-10-15

    Highlights: • We summarized most of the challenges of fusion devices to robot systems. • Propose an active cooling system to protect all of the necessary components. • Trial design test and theoretical analysis were conducted. • Overall implementation of the active cooling system was demonstrated. - Abstract: In-vessel operation/inspection is an indispensable task for Tokamak experimental reactor, for a robot/manipulator is more capable in doing this than human being with more precise motion and less risk of damaging the ambient equipment. Considering the demanding conditions of Tokamak, the manipulator should be adaptable to rapid response in the extreme conditions such as high temperature, vacuum and so on. In this paper, we propose an active cooling system embedded into such manipulator. Cameras, motors, gearboxes, sensors, and other mechanical/electrical components could then be designed under ordinary conditions. The cooling system cannot only be a thermal shield since the components are also heat sources in dynamics. We carry out a trial test to verify our proposal, and analyze the active cooling system theoretically, which gives a direction on the optimization by varying design parameters, components and distribution. And based on thermal sensors monitoring and water flow adjusting a closed-loop feedback control of temperature is added to the system. With the preliminary results, we believe that the proposal gives a way to robust and inexpensive design in extreme environment. Further work will concentrate on overall implementation and evaluation of this cooling system with the whole inspection manipulator.

  12. Evaluation of Active Cooling Systems for Non-Residential Buildings

    Directory of Open Access Journals (Sweden)

    M.A. Othuman Mydin

    2014-05-01

    Full Text Available Cooling systems are an essential element in many facets of modern society including cars, computers and buildings. Cooling systems are usually divided into two types: passive and active. Passive cooling transfers heat without using any additional energy while active cooling is a type of heat transfer that uses powered devices such as fans or pumps. This paper will focus on one particular type of passive cooling: air-conditioning systems. An air-conditioning system is defined as controlled air movement, temperature, humidity and cleanliness of a building area. Air conditioning consists of cooling and heating. Therefore, the air-conditioning system should be able to add and remove heat from the area. An air-conditioning system is defined as a control or treatment of air in a confined space. The process that occurs is the air-conditioning system absorbs heat and dust while, at the same time, cleaning the air breathed into a closed space. The purpose of air-conditioning is to maintain a comfortable atmosphere for human life and to meet user requirements. In this paper, air-conditioning systems for non-residential buildings will be presented and discussed.

  13. Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

    International Nuclear Information System (INIS)

    Althaus; Jansing; Kesseler; Kirchner; Menck

    1974-01-01

    Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

  14. A model for radionuclide transport in the Cooling Water System

    International Nuclear Information System (INIS)

    Kahook, S.D.

    1992-08-01

    A radionuclide transport model developed to assess radiological levels in the K-reactor Cooling Water System (CWS) in the event of an inadvertent process water (PW) leakage to the cooling water (CW) in the heat exchangers (HX) is described. During and following a process water leak, the radionuclide transport model determines the time-dependent release rates of radionuclide from the cooling water system to the environment via evaporation to the atmosphere and blow-down to the Savannah River. The developed model allows for delay times associated with the transport of the cooling water radioactivity through cooling water system components. Additionally, this model simulates the time-dependent behavior of radionuclides levels in various CWS components. The developed model is incorporated into the K-reactor Cooling Tower Activity (KCTA) code. KCTA allows the accident (heat exchanger leak rate) and the cooling tower blow-down and evaporation rates to be described as time-dependent functions. Thus, the postulated leak and the consequence of the assumed leak can be modelled realistically. This model is the first of three models to be ultimately assembled to form a comprehensive Liquid Pathway Activity System (LPAS). LPAS will offer integrated formation, transport, deposition, and release estimates for radionuclides formed in a SRS facility. Process water and river water modules are forthcoming as input and downstream components, respectively, for KCTA

  15. Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Araj, K.

    1983-01-01

    The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

  16. Fast Flux Test Facility core system

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

    1990-11-01

    A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

  17. Post-LOCA core flushing system

    International Nuclear Information System (INIS)

    Boyajian, J.D.; Weinberger, P.A.

    1980-01-01

    A system is disclosed for flushing the core of a nuclear reactor after a loss-of-coolant accident. A pump causes flow of liquid-phase fluid from the containment-vessel sump. This flow is used to provide the motivating force for an eductor that causes suction at the hot log of the reactor. The eductor suction can draw gas-phase coolant out of the hot leg. As a result, it can reduce pressure which may be preventing the flow of liquid-phase coolant out of the hot leg. By causing liquid-phase flow through the reactor, the system ensures that particles and boric acid are flushed out of the core. The system thereby eliminates the build-up of particles and the concentrations of boric acid in the core that could result if the coolant were to leave the pressure vessel exclusively in the gas phase. 9 claims

  18. System and method for pre-cooling of buildings

    Science.gov (United States)

    Springer, David A.; Rainer, Leo I.

    2011-08-09

    A method for nighttime pre-cooling of a building comprising inputting one or more user settings, lowering the indoor temperature reading of the building during nighttime by operating an outside air ventilation system followed, if necessary, by a vapor compression cooling system. The method provides for nighttime pre-cooling of a building that maintains indoor temperatures within a comfort range based on the user input settings, calculated operational settings, and predictions of indoor and outdoor temperature trends for a future period of time such as the next day.

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

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

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

  20. Hybrid Cooling System for Industrial Application | Ezekwe | Nigerian ...

    African Journals Online (AJOL)

    Hybrid Cooling System for Industrial Application. ... PROMOTING ACCESS TO AFRICAN RESEARCH ... more than five times over that achieved by using the gas (air) phase alone. ... EMAIL FREE FULL TEXT EMAIL FREE FULL TEXT

  1. Augmented cooling vest system subassembly: Design and analysis

    International Nuclear Information System (INIS)

    D’Angelo, Maurissa; D’Angelo, Joseph; Almajali, Mohammad; Lafdi, Khalid; Delort, Antoine; Elmansori, Mohamed

    2014-01-01

    Highlights: • Thermoelectric cooler (TEC) was employed to provide cooling air to cooling vest. • Aluminum cooling fins were used to exchange heat for hot and cold sides of TEC. • Performance of the system was determined and the experimental technique was described. • Heat sink is capable to remove additional heat and heat exchanger provides cooling air. • Future work is proposed to optimize the efficiency of the system. - Abstract: A prototype cooling engine consisting of thermoelectric coolers (TECs) was developed and designed. In this prototype, aluminum cooling fins were employed as the heat exchange method for both the hot and cold sides of the TEC. Aluminum fins were used to cool the ambient air through a heat exchanger and dissipate heat build up from the heat sink. This system was modeled and performance capabilities were determined. The experimental technique used to monitor parameters affecting the efficiency of the designed system was described. These parameters include the temperatures of the inlets and outlets of both heat exchanger and heat sink and the flow rate of the cooled air. The experiment was run under three input DC powers; 15 V, 18 V, and 21 V. As the power increased, both the flow rate and the temperature difference between the hot and cold side of thermoelectric cooler increased, demonstrating the heat sink capability to remove the additional heat. However, the temperature difference between the inlet and outlet of the heat exchanger decreases as the power increase. The findings demonstrated the effectiveness of this cooling system and future work is proposed to optimize the heat

  2. New cooling system of the FRG-1 two barrier system of the primary coolant cycle

    International Nuclear Information System (INIS)

    Knop, W.; Schreiner, P.

    2003-01-01

    The GKSS research center operates the swimming pool reactor FRG-1 with a thermal power of 5 MW as national neutron source for neutron scattering experiments and sample irradiation as well. Before changing the primary coolant cycle consisted of the reactor core and the closed piping including pumps, heat exchanger and delay tank. The closed cooling circuit was located underneath the reactor pool, in the so-called radioactive cellar. This piping system served secondary coolant system. Due to the location of the primary coolant cycle below the operation pool a postulated 2-F line break and simultaneous failure of the pool slide gate valve could lead to a falling dry of the total reactor core. the new primary coolant system was built in the beginning 2002 in a partitioned cell all within the radioactive cellar, so that the reactor core remains with water with the assumed incident. Due to the new two barrier-inclusion of the primary circuit only the melting of two fuel plates (from total 252 fuel plates) has to be taken into account. This measure and the core compactness in 2000 with a neutron flux gain of a factor of 2 makes the FRG-1 ready for the next 15 years of reactor operation. (author)

  3. Upgrade of the cooling water temperature measures system for HLS

    International Nuclear Information System (INIS)

    Guo Weiqun; Liu Gongfa; Bao Xun; Jiang Siyuan; Li Weimin; He Duohui

    2007-01-01

    The cooling water temperature measures system for HLS (Hefei Light Source) adopts EPICS to the developing platform and takes the intelligence temperature cruise instrument for the front control instrument. Data of temperatures are required by IOCs through Serial Port Communication, archived and searched by Channel Archiver. The system can monitor the real-time temperatures of many channels cooling water and has the function of history data storage, and data network search. (authors)

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

    International Nuclear Information System (INIS)

    EL Khatib, H.H.A.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Utaja

    2002-01-01

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

  6. Process integration: Cooling water systems design

    CSIR Research Space (South Africa)

    Gololo, KV

    2010-10-01

    Full Text Available stream_source_info Gololo2_2010.pdf.txt stream_content_type text/plain stream_size 17891 Content-Encoding UTF-8 stream_name Gololo2_2010.pdf.txt Content-Type text/plain; charset=UTF-8 The 13th Asia Pacific Confederation... results in a nonlinear program (NLP) formulation and the second case yields mixed integer nonlinear program (MINLP). In both cases the cooling towers operating capacity were debottlenecked without compromising the heat duties. The 13th Asia...

  7. Water cooling system for sintering furnaces of nuclear fuel pellets

    International Nuclear Information System (INIS)

    1996-01-01

    This work has as a main objective to develop a continuous cooling water system, which is necessary for the cooling of the sintering furnaces. This system is used to protect them as well as for reducing the water consumption, ejecting the heat generated into this furnaces and scattering it into the atmosphere in a fast and continuous way. The problem was defined and the reference parameters established, making the adequate research. The materials were selected as well as the length of the pipeline which will carry the secondary refrigerant fluid (water). Three possible solutions were tried,and evaluated, and from these, the thermal and economically most efficient option was selected. The layout of the solution was established and the theoretical construction of a cooling system for liquids using dichlorofluoromethane (R-22), as a refrigerant and a air cooled condenser, was accomplished. (Author)

  8. Adsorption Cooling System Using Metal-Impregnated Zeolite-4A

    Directory of Open Access Journals (Sweden)

    Somsuk Trisupakitti

    2016-01-01

    Full Text Available The adsorption cooling systems have been developed to replace vapor compression due to their benefits of being environmentally friendly and energy saving. We prepared zeolite-4A and experimental cooling performance test of zeolite-water adsorption system. The adsorption cooling test-rig includes adsorber, evaporator, and condenser which perform in vacuum atmosphere. The maximum and minimum water adsorption capacity of different zeolites and COP were used to assess the performance of the adsorption cooling system. We found that loading zeolite-4A with higher levels of silver and copper increased COP. The Cu6%/zeolite-4A had the highest COP at 0.56 while COP of zeolite-4A alone was 0.38. Calculating the acceleration rate of zeolite-4A when adding 6% of copper would accelerate the COP at 46%.

  9. Performance comparison between a solar driven rotary desiccant cooling system and conventional vapor compression system (performance study of desiccant cooling)

    International Nuclear Information System (INIS)

    Ge, T.S.; Ziegler, F.; Wang, R.Z.; Wang, H.

    2010-01-01

    Solar driven rotary desiccant cooling systems have been widely recognized as alternatives to conventional vapor compression systems for their merits of energy-saving and being eco-friendly. In the previous paper, the basic performance features of desiccant wheel have been discussed. In this paper, a solar driven two-stage rotary desiccant cooling system and a vapor compression system are simulated to provide cooling for one floor in a commercial office building in two cities with different climates: Berlin and Shanghai. The model developed in the previous paper is adopted to predict the performance of the desiccant wheel. The objectives of this paper are to evaluate and compare the thermodynamic and economic performance of the two systems and to obtain useful data for practical application. Results show that the desiccant cooling system is able to meet the cooling demand and provide comfortable supply air in both of the two regions. The required regeneration temperatures are 55 deg. C in Berlin and 85 deg. C in Shanghai. As compared to the vapor compression system, the desiccant cooling system has better supply air quality and consumes less electricity. The results of the economic analysis demonstrate that the dynamic investment payback periods are 4.7 years in Berlin and 7.2 years in Shanghai.

  10. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

    2013-01-01

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

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

    Science.gov (United States)

    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).

  12. Development of adsorption cooling system. 3

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.N.; Cho, S.H.; Chue, K.T.; You, Y.J.; Lee, K.H.; Eun, T.H. [Korea Inst. of Energy Research, Taejon (Korea, Republic of)

    1995-12-01

    This report describes the third year study to develop adsorption chiller using silica gel/water pair for the recovery of low level waste heat. A pilot plant was fabricated and tested. In a typical run, the cooling capacity of 1.66 USRT and COP of 0.38 was obtained under the following operating conditions; chilled water temperature of 12{yields}8.9 degree C, hot water temperature of 72.7 degree C, cooling water temperature of 23.2 degree C, and half cycle time of 600(s). The COP of the pilot plant is comparable to 0.4 of Nishiyodo pilot plant having 3.68 USRT. In order to enhance the thermal conductivity of adsorbent layer, consolidated silica gel and graphite block was prepared and its characteristics was analyzed. A slurry method using water was appropriate of silica gel and graphite in the block, in which adsorbed amount of water is not much smaller than that on silica gel, was 6:1. The thermal conductivity of this block was 6.53 W/mk which was 37 times larger than that of silica gel. (author). 12 refs., 37 figs., 8 tabs.

  13. Simulation of an active cooling system for photovoltaic modules

    International Nuclear Information System (INIS)

    Abdelhakim, Lotfi

    2016-01-01

    Photovoltaic cells are devices that convert solar radiation directly into electricity. However, solar radiation increases the photovoltaic cells temperature [1] [2]. The temperature has an influence on the degradation of the cell efficiency and the lifetime of a PV cell. This work reports on a water cooling technique for photovoltaic panel, whereby the cooling system was placed at the front surface of the cells to dissipate excess heat away and to block unwanted radiation. By using water as a cooling medium for the photovoltaic solar cells, the overheating of closed panel is greatly reduced without prejudicing luminosity. The water also acts as a filter to remove a portion of solar spectrum in the infrared band but allows transmission of the visible spectrum most useful for the PV operation. To improve the cooling system efficiency and electrical efficiency, uniform flow rate among the cooling system is required to ensure uniform distribution of the operating temperature of the PV cells. The aims of this study are to develop a 3D thermal model to simulate the cooling and heat transfer in Photovoltaic panel and to recommend a cooling technique for the PV panel. The velocity, pressure and temperature distribution of the three-dimensional flow across the cooling block were determined using the commercial package, Fluent. The second objective of this work is to study the influence of the geometrical dimensions of the panel, water mass flow rate and water inlet temperature on the flow distribution and the solar panel temperature. The results obtained by the model are compared with experimental results from testing the prototype of the cooling device.

  14. Simulation of an active cooling system for photovoltaic modules

    Energy Technology Data Exchange (ETDEWEB)

    Abdelhakim, Lotfi [Széchenyi István University of Applied Sciences, Department of Mathematics, P.O.Box 701, H-9007 Győr (Hungary)

    2016-06-08

    Photovoltaic cells are devices that convert solar radiation directly into electricity. However, solar radiation increases the photovoltaic cells temperature [1] [2]. The temperature has an influence on the degradation of the cell efficiency and the lifetime of a PV cell. This work reports on a water cooling technique for photovoltaic panel, whereby the cooling system was placed at the front surface of the cells to dissipate excess heat away and to block unwanted radiation. By using water as a cooling medium for the photovoltaic solar cells, the overheating of closed panel is greatly reduced without prejudicing luminosity. The water also acts as a filter to remove a portion of solar spectrum in the infrared band but allows transmission of the visible spectrum most useful for the PV operation. To improve the cooling system efficiency and electrical efficiency, uniform flow rate among the cooling system is required to ensure uniform distribution of the operating temperature of the PV cells. The aims of this study are to develop a 3D thermal model to simulate the cooling and heat transfer in Photovoltaic panel and to recommend a cooling technique for the PV panel. The velocity, pressure and temperature distribution of the three-dimensional flow across the cooling block were determined using the commercial package, Fluent. The second objective of this work is to study the influence of the geometrical dimensions of the panel, water mass flow rate and water inlet temperature on the flow distribution and the solar panel temperature. The results obtained by the model are compared with experimental results from testing the prototype of the cooling device.

  15. Emergency reactor cooling device

    International Nuclear Information System (INIS)

    Arakawa, Ken.

    1993-01-01

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

  16. Active noise canceling system for mechanically cooled germanium radiation detectors

    Science.gov (United States)

    Nelson, Karl Einar; Burks, Morgan T

    2014-04-22

    A microphonics noise cancellation system and method for improving the energy resolution for mechanically cooled high-purity Germanium (HPGe) detector systems. A classical adaptive noise canceling digital processing system using an adaptive predictor is used in an MCA to attenuate the microphonics noise source making the system more deployable.

  17. System performance and economic analysis of solar-assisted cooling/heating system

    KAUST Repository

    Huang, B.J.; Wu, J.H.; Yen, R.H.; Wang, J.H.; Hsu, H.Y.; Hsia, C.J.; Yen, C.W.; Chang, J.M.

    2011-01-01

    The long-term system simulation and economic analysis of solar-assisted cooling/heating system (SACH-2) was carried out in order to find an economical design. The solar heat driven ejector cooling system (ECS) is used to provide part of the cooling

  18. A parametric study of solar operated cooling system

    International Nuclear Information System (INIS)

    Zagalei, Abdullatif Salin

    2006-01-01

    Because of energy for air conditioning has been the fastest-growing segment of energy of consumption market in Libya and generally in north Africa, and with the realization depleting nature of fossil fuel, solar cooling of buildings which leads to the improvement of human comfort represents a potentially significant application of solar energy where the availability of solar radiation meets with the cooling load demand. This application has been shown to be technically feasible but the equipment needs further investigative research to improve its performance and feasibility. A solar operated absorption cooling system with energy storage is selected. A latent heat storage would be a space saver for such application for solar energy. A system modeling is an essential activity in order to go for system simulation. A complete solar cooling system to be modeled through the thermodynamic analysis of each system components. Resulting a package of equations used directly to the system simulation in order to predict the system performance to obtain the optimum working conditions for the selected cooling system. A computer code which is used to simulate a series of calculations was written in Fortran language according to the constructed information flow diagram and simulation program flow char. For a typical input data a set of results are reported and discussed and shows that the selected system promises to be a good choice for air conditioning application in Libya specially for large building as storehouses, shopping centers, public administrative.(Author)

  19. Comparative Studies of Core Thermal Hydraulic Design Methods for the Prototype Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Lim, Jae Yong; Kim, Sang Ji

    2013-01-01

    In this work, various core thermal-hydraulic design methods, which have arisen during the development of a prototype SFR, are compared to establish a proper design procedure. Comparative studies have been performed to determine the appropriate design method for the prototype SFR. The results show that the minimization method show a lower cladding midwall temperature than the fixed outlet temperature methods and superior thermal safety margin with the same coolant flow. The Korea Atomic energy Research Institute (KAERI) has performed a conceptual SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damages in SFR subassemblies are arisen from a creep induced failure. The creep limit is evaluated based on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, the core thermalhydraulic design method, which eventually determines the cladding temperature, is highly important to assure a safe and reliable operation of the reactor systems

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

    International Nuclear Information System (INIS)

    Chen, Zhao; Zhao, Pengcheng; Zhou, Guangming; Chen, Hongli

    2014-01-01

    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

  1. Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

  2. Thermohydraulic modeling of the dry air passive containment cooling system process in the Westinghouse AP-600 ALWR

    Energy Technology Data Exchange (ETDEWEB)

    Harari, R; Weis, Y; Barnea, Y [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    Following postulated events of a LOCA, the passive Containment Cooling System (PCCS) uses dry air to transfer the residual heat by natural circulation. The air flow path, designed between the steel reactor containment hot shell and the concrete shield building, creates an open thermosyphon. The purpose of this inherently safe process is to assure the long term steady-state cooling of the nuclear core after an emergency shutdown (authors).

  3. Air-cooled recirculation cooling systems. Technical and economic comparison; Luftgekuehlte Rueckkuehlsysteme. Technisch wirtschaftlicher Vergleich

    Energy Technology Data Exchange (ETDEWEB)

    Dierks, G. [Fa. Jaeggi/Guentner (Schweiz) AG, Trimbach (Switzerland)

    2000-03-01

    There are several air-cooled forced-circulation cooling systems for heat removal from refrigeration systems. Optimum solutions should not be selected on the basis of the cost factor alone; an integrative approach should be used instead. An exemplary investigation is presented. [German] Fuer die Waermeabfuhr aus kaeltetechnischen Anlagen stehen verschiedene luftgekuehlte, zwangsbelueftete Rueckkuehlsysteme zur Verfuegung. Die Auswahl des Systems ist oft von kurzfristigem Kostendenken gepraegt, was in technischer und wirtschaftlicher Hinsicht aber nicht immer der optimalen Loesung entspricht. Erst die genauere Kenntnis der verschiedenen Systeme und eine ganzheitliche Betrachtungsweise ermoeglichen die optimale Wahl fuer den einzelnen Fall. Die hier praesentierte Untersuchung wird anhand eines konkreten Falls dargestellt, wobei Preise und technische Produktdaten auf realen Anfragen beruhen. Der Autor ist um objetive Bewertung bemueht, der Leser moege aber selbst urteilen. (orig./AKF)

  4. Passive ventilation systems with heat recovery and night cooling

    DEFF Research Database (Denmark)

    Hviid, Christian Anker; Svendsen, Svend

    2008-01-01

    with little energy consumption and with satisfying indoor climate. The concept is based on using passive measures like stack and wind driven ventilation, effective night cooling and low pressure loss heat recovery using two fluid coupled water-to-air heat exchangers developed at the Technical University......In building design the requirements for energy consumption for ventilation, heating and cooling and the requirements for increasingly better indoor climate are two opposing factors. This paper presents the schematic layout and simulation results of an innovative multifunc-tional ventilation concept...... of Denmark. Through building integration in high performance offices the system is optimized to incorporate multiple functions like heating, cooling and ventilation, thus saving the expenses of separate cooling and heating systems. The simulation results are derived using the state-of-the-art building...

  5. Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Scheuerer, Martina, E-mail: Martina.Scheuerer@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany); Weis, Johannes, E-mail: Johannes.Weis@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Pressurized thermal shocks are important phenomena for plant life extension and aging. Black-Right-Pointing-Pointer The thermal-hydraulics of PTS have been studied experimentally and numerically. Black-Right-Pointing-Pointer In the Large Scale Test Facility a loss of coolant accident was investigated. Black-Right-Pointing-Pointer CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

  6. Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

    International Nuclear Information System (INIS)

    Scheuerer, Martina; Weis, Johannes

    2012-01-01

    Highlights: ► Pressurized thermal shocks are important phenomena for plant life extension and aging. ► The thermal-hydraulics of PTS have been studied experimentally and numerically. ► In the Large Scale Test Facility a loss of coolant accident was investigated. ► CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

  7. Instrumentation for NBI SST-1 cooling water system

    International Nuclear Information System (INIS)

    Qureshi, Karishma; Patel, Paresh; Jana, M.R.

    2015-01-01

    Neutral Beam Injector (NBI) System is one of the heating systems for Steady state Superconducting Tokamak (SST-1). It is capable of generating a neutral hydrogen beam of power 0.5 MW at 30 kV. NBI system consists of following sub-systems: Ion source, Neutralizer, Deflection Magnet and Magnet Liner (ML), Ion Dump (ID), V-Target (VT), Pre Duct Scraper (PDS), Beam Transmission Duct (BTD) and Shine Through (ST). For better heat removal management purpose all the above sub-systems shall be equipped with Heat Transfer Elements (THE). During beam operation these sub-systems gets heated due to the received heat load which requires to be removed by efficient supplying water. The cooling water system along with the other systems (External Vacuum System, Gas Feed System, Cryogenics System, etc.) will be controlled by NBI Programmable Logic Control (PLC). In this paper instrumentation and its related design for cooling water system is discussed. The work involves flow control valves, transmitters (pressure, temperature and water flow), pH and conductivity meter signals and its interface with the NBI PLC. All the analog input, analog output, digital input and digital output signals from the cooling water system will be isolated and then fed to the NBI PLC. Graphical Users Interface (GUI) needed in the Wonderware SCADA for the cooling water system shall also be discussed. (author)

  8. Supplementary report: cooling water systems for Darlington G.S

    International Nuclear Information System (INIS)

    1975-08-01

    This report summarizes Ontario Hydro's existing aquatic environmental programs, presents results of these investigations, and outlines plans and activities for expanded aquatic environment studies including the evaluation of alternative cooling systems. This report outlines specific considerations regarding possible alternative cooling arrangements for the Darlington station. It concludes with a recommendation that a study be initiated to examine the potential benefits of using the heated discharge water in a warm water recreational centre. (author)

  9. Finite impulse testing (FIT) system for Emergency Cooling System (ECS) in Dhruva

    International Nuclear Information System (INIS)

    Punekar, Parag; Ramkumar, N.; Kulkarni, U.S.; Darbhea, M.D.; Bharadhwaj, G; Jangra, L.R.; Geetha, Patil; Das, Shantanu; Sonnis, S.T.; Trevedi, P.; Patil, M.B.; Biswas, B.B.

    2006-01-01

    Finite Impulse Testing (FIT) system for Emergency Cooling System (ECS) is used to check healthiness of ECS logic circuits in an online mode. The ECS is an important safety system that ensures the cooling of reactor core during shutdown state of Main Coolant Pumps (MCPs), and hence FIT-ECS that monitors the health of ECS logic circuits in an online (real time) mode is an important part of it. Based on a Safety Related Unusual Occurrence in ECS system due to the malfunction of its earlier single channel FIT system, the new FIT-ECS system has been designed with new features and is commissioned. The FIT-ECS system feeds the simulated input signals (fine impulses of nominal width 575 μS) to the ECS logic circuits and read the outputs. These output (predicted) signals from ECS logic circuit are processed in the FIT-ECS system and in event of any discrepancy, the FIT-ECS system displays fault signature on local panel, detailed information of the fault on Operator Console (OC), and generates an alarm 'ECS logic Fail' in the control room. FIT-ECS also monitors the inputs and outputs of ECS logic circuit. All the information required is stored as a database that can be subsequently displayed in various formats. ECS system is designated as Category I-A system and is a hardwired system and FIT-ECS monitors the healthiness of the logics of the ECS System is a computerized system. As per IEC 1226, FIT-ECS is categorized as Category I-B system. This paper provides technical information on FIT-ECS system design, its important features, the testing carried on the FIT-ECS system, interconnections of FIT-ECS and ECS and the commissioning experience of FIT-ECS system. (author)

  10. System Design of a Supercritical CO_2 cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

  11. Uncovering the information core in recommender systems

    Science.gov (United States)

    Zeng, Wei; Zeng, An; Liu, Hao; Shang, Ming-Sheng; Zhou, Tao

    2014-08-01

    With the rapid growth of the Internet and overwhelming amount of information that people are confronted with, recommender systems have been developed to effectively support users' decision-making process in online systems. So far, much attention has been paid to designing new recommendation algorithms and improving existent ones. However, few works considered the different contributions from different users to the performance of a recommender system. Such studies can help us improve the recommendation efficiency by excluding irrelevant users. In this paper, we argue that in each online system there exists a group of core users who carry most of the information for recommendation. With them, the recommender systems can already generate satisfactory recommendation. Our core user extraction method enables the recommender systems to achieve 90% of the accuracy of the top-L recommendation by taking only 20% of the users into account. A detailed investigation reveals that these core users are not necessarily the large-degree users. Moreover, they tend to select high quality objects and their selections are well diversified.

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

    International Nuclear Information System (INIS)

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

    1994-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-04-01

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

  14. Mapping the particle acceleration in the cool core of the galaxy cluster RX J1720.1+2638

    International Nuclear Information System (INIS)

    Giacintucci, S.; Markevitch, M.; Brunetti, G.; Venturi, T.; ZuHone, J. A.; Mazzotta, P.; Bourdin, H.

    2014-01-01

    We present new deep, high-resolution radio images of the diffuse minihalo in the cool core of the galaxy cluster RX J1720.1+2638. The images have been obtained with the Giant Metrewave Radio Telescope at 317, 617, and 1280 MHz and with the Very Large Array at 1.5, 4.9, and 8.4 GHz, with angular resolutions ranging from 1'' to 10''. This represents the best radio spectral and imaging data set for any minihalo. Most of the radio flux of the minihalo arises from a bright central component with a maximum radius of ∼80 kpc. A fainter tail of emission extends out from the central component to form a spiral-shaped structure with a length of ∼230 kpc, seen at frequencies 1.5 GHz and below. We find indication of a possible steepening of the total radio spectrum of the minihalo at high frequencies. Furthermore, a spectral index image shows that the spectrum of the diffuse emission steepens with increasing distance along the tail. A striking spatial correlation is observed between the minihalo emission and two cold fronts visible in the Chandra X-ray image of this cool core. These cold fronts confine the minihalo, as also seen in numerical simulations of minihalo formation by sloshing-induced turbulence. All these observations favor the hypothesis that the radio-emitting electrons in cluster cool cores are produced by turbulent re-acceleration.

  15. Core cooling and thermal responses during whole-head, facial, and dorsal immersion in 17 degrees C water.

    Science.gov (United States)

    Pretorius, Thea; Gagnon, Dominique D; Giesbrecht, Gordon G

    2010-10-01

    This study isolated the effects of dorsal, facial, and whole-head immersion in 17 degrees C water on peripheral vasoconstriction and the rate of body core cooling. Seven male subjects were studied in thermoneutral air (approximately 28 degrees C). On 3 separate days, they lay prone or supine on a bed with their heads inserted through the side of an adjustable immersion tank. Following 10 min of baseline measurements, the water level was raised such that the water immersed the dorsum, face, or whole head, with the immersion period lasting 60 min. During the first 30 min, the core (esophageal) cooling rate increased from dorsum (0.29 ± 0.2 degrees C h-1) to face (0.47 ± 0.1 degrees C h-1) to whole head (0.69 ± 0.2 degrees C h(-1)) (p whole-head immersion (114 ± 52% h(-1)) than in either facial (51 ± 47% h-1) or dorsal (41 ± 55% h(-1)) immersion (p whole-head (120.5 ± 13 kJ), facial (86.8 ± 17 kJ), and dorsal (46.0 ± 11 kJ) immersion (p whole head elicited a higher rate of vasoconstriction, the face did not elicit more vasoconstriction than the dorsum. Rather, the progressive increase in core cooling from dorsal to facial to whole-head immersion simply correlates with increased heat loss.

  16. Dry-type cooling systems in electric power production

    International Nuclear Information System (INIS)

    Li, K.W.

    1973-01-01

    This study indicates that the dry-type cooling tower could be adopted in this country as an alternative method for removing waste heat from power plants. The use of dry cooling towers would not only lead to a change of cooling system design, but also to a change of overall thermal design in a power generating system. The principal drawbacks to using dry cooling towers in a large steam-turbine plant are the generating capacity loss, increased fuel consumption and the high capital cost of the dry cooling towers. These economic penalties must be evaluated in each specific case against the benefits that may result from the use of dry cooling towers. The benefits are principally these: (1) Fewer constraints in the selection of power plant sites, (2) No thermal discharge to the natural water bodies, (3) Elimination of vapor plumes and water evaporation loss, and (4) Freedom of adding new units to an existing facility where inadequate water supply may otherwise rule out this possibility

  17. Replacement inhibitors for tank farm cooling coil systems

    International Nuclear Information System (INIS)

    Hsu, T.C.

    1995-01-01

    Sodium chromate has been an effective corrosion inhibitor for the cooling coil systems in Savannah River Site (SRS) waste tanks for over 40 years. Due to their age and operating history, cooling coils occasionally fail allowing chromate water to leak into the environment. When the leaks spill 10 lbs. or more of sodium chromate over a 24-hr period, the leak incidents are classified as Unusual Occurrences (UO) per CERCLA (Comprehensive Environmental Response, Compensation and Liability Act). The cost of reporting and cleaning up chromate spills prompted High Level Waste Engineering (HLWE) to initiate a study to investigate alternative tank cooling water inhibitor systems and the associated cost of replacement. Several inhibitor systems were investigated as potential alternatives to sodium chromate. All would have a lesser regulatory impact, if a spill occurred. However, the conversion cost is estimated to be $8.5 million over a period of 8 to 12 months to convert all 5 cooling systems. Although each of the alternative inhibitors examined is effective in preventing corrosion, there is no inhibitor identified that is as effective as chromate. Assuming 3 major leaks a year (the average over the past several years), the cost of maintaining the existing inhibitor was estimated at $0.5 million per year. Since there is no economic or regulatory incentive to replace the sodium chromate with an alternate inhibitor, HLWE recommends that sodium chromate continue to be used as the inhibitor for the waste tank cooling systems

  18. Development of a higher power cooling system for lithium targets.

    Science.gov (United States)

    Phoenix, B; Green, S; Scott, M C; Bennett, J R J; Edgecock, T R

    2015-12-01

    The accelerator based Boron Neutron Capture Therapy beam at the University of Birmingham is based around a solid thick lithium target cooled by heavy water. Significant upgrades to Birmingham's Dynamitron accelerator are planned prior to commencing a clinical trial. These upgrades will result in an increase in maximum achievable beam current to at least 3 mA. Various upgrades to the target cooling system to cope with this increased power have been investigated. Tests of a phase change coolant known as "binary ice" have been carried out using an induction heater to provide a comparable power input to the Dynamitron beam. The experimental data shows no improvement over chilled water in the submerged jet system, with both systems exhibiting the same heat input to target temperature relation for a given flow rate. The relationship between the cooling circuit pumping rate and the target temperature in the submerged jet system has also been tested. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Development and application of online Stelmor Controlled Cooling System

    International Nuclear Information System (INIS)

    Yu Wanhua; Chen Shaohui; Kuang Yonghai; Cao Kaichao

    2009-01-01

    An online Stelmor Controlled Cooling System (SCCS) has been developed successfully for the Stelmor production line, which can communicate with the material flow management system and Program Logic Control System (PLCs) automatically through local network. This online model adopts Implicit Finite Difference Time Domain (FDTD) method to calculate temperature evolution and phase transformation during the production process and predicts final properties. As Continuous Cooling Temperature (CCT) curves of various steels can be coupled in the model, it can predict the latent heat rise and range of phase transformation for various steels, which can provide direct guidance for new steel development and optimization of present Stelmor cooling process. This unique online system has been installed in three Stelmor production lines at present with good results.

  20. 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)

    2001-02-01

    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

  1. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  2. Cooling system upgrading from 250 kW to 1 MW

    International Nuclear Information System (INIS)

    Anderson, T.V.; Johnson, A.G.; Ringle, J.C.

    1972-01-01

    The Oregon State TRIGA reactor (OSTR) power capability was upgraded from 250 KW to 1 MW in 1969; however, funds were not available for simultaneous upgrading of the cooling system. Since then, the OSTR has been selectively operating at full power with the original 250 KW cooling system. After funds were made available in 1971 the construction on the new heat exchanger building began. The new cooling system was installed, equipment was checked out, corrections were made, and acceptance tests were run. In addition, several days were required to clean up the primary system water, since increased water flow (350 gpm) swirled 4 year's collection of sediment off the reactor tank bottom and into the primary system. Three interesting items have been noticed, which are apparently a result of the cooling system upgrading: (1) the radiation levels above the reactor tank have been reduced by a factor of 2 to 3, (2) a low resonance vibration in the reactor core occurs at 1 MW. The vibration is attributed to a combination of increased water turbulence and subcooled (surface) nucleate boiling, and (3) direct radiation levels from the demineralizer tank have increased approximately 8-fold. This resulted in a relocation of the tank and the use of supplemental shielding. Increased operating time at higher average power levels, plus disturbance of; sediment on the bottom of the reactor tank are believed to be the main sources of the higher radiation levels

  3. Control of Non-linear Marine Cooling System

    DEFF Research Database (Denmark)

    Hansen, Michael; Stoustrup, Jakob; Bendtsen, Jan Dimon

    2011-01-01

    We consider the problem of designing control laws for a marine cooling system used for cooling the main engine and auxiliary components aboard several classes of container vessels. We focus on achieving simple set point control for the system and do not consider compensation of the non-linearitie......-linearities, closed circuit flow dynamics or transport delays that are present in the system. Control laws are therefore designed using classical control theory and the performance of the design is illustrated through two simulation examples....

  4. COMMIX analysis of AP-600 Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Chang, J.F.C.; Chien, T.H.; Ding, J.; Sun, J.G.; Sha, W.T.

    1992-01-01

    COMMIX modeling and basic concepts that relate components, i.e., containment, water film cooling, and natural draft air flow systems. of the AP-600 Passive Containment Cooling System are discussed. The critical safety issues during a postulated accident have been identified as (1) maintaining the liquid film outside the steel containment vessel, (2) ensuring the natural convection in the air annulus. and (3) quantifying both heat and mass transfer accurately for the system. The lack of appropriate heat and mass transfer models in the present analysis is addressed. and additional assessment and validation of the proposed models is proposed

  5. System for cooling hybrid vehicle electronics, method for cooling hybrid vehicle electronics

    Science.gov (United States)

    France, David M.; Yu, Wenhua; Singh, Dileep; Zhao, Weihuan

    2017-11-21

    The invention provides a single radiator cooling system for use in hybrid electric vehicles, the system comprising a surface in thermal communication with electronics, and subcooled boiling fluid contacting the surface. The invention also provides a single radiator method for simultaneously cooling electronics and an internal combustion engine in a hybrid electric vehicle, the method comprising separating a coolant fluid into a first portion and a second portion; directing the first portion to the electronics and the second portion to the internal combustion engine for a time sufficient to maintain the temperature of the electronics at or below 175.degree. C.; combining the first and second portion to reestablish the coolant fluid; and treating the reestablished coolant fluid to the single radiator for a time sufficient to decrease the temperature of the reestablished coolant fluid to the temperature it had before separation.

  6. analysis of reactivity accidents in MTR for various protection system parameters and core condition

    International Nuclear Information System (INIS)

    Mohamed, F.M.

    2011-01-01

    Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myoung Jun; Lee, Hee Joon [Kookmin University, Seoul (Korea, Republic of); Moon, Joo Hyung; Bae, Youngmin; Kim, Young-In [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

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

  8. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

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

    International Nuclear Information System (INIS)

    Lacerda, F.C.

    1987-01-01

    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) [pt

  10. Safety analysis of reactor's cooling system

    International Nuclear Information System (INIS)

    1999-01-01

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

  11. Reactor core flow rate control system

    International Nuclear Information System (INIS)

    Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

    1996-01-01

    When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

  12. Overview of core designs and requirements/criteria for core restraint systems

    International Nuclear Information System (INIS)

    Sutherland, W.H.

    1984-09-01

    The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

  13. SCORPIO: a framework for core surveillance systems

    International Nuclear Information System (INIS)

    Berg, Oe.; Tsuiki, Makoto

    1999-01-01

    The first version of the core surveillance system SCORPIO was installed at Unit 2, Ringhals, in 1984. It was implemented on Norsk Data mini-computers with a fully graphical user-interface. The main purpose was to provide a practical tool for reactor operators and reactor physicists for on-line monitoring and predictive analysis of core behaviour. A second version of SCORPIO was developed in 1993-1995 and implemented on Unix workstations. In addition to upgrading the system at Ringhals, the system was installed by Duke Power, USA, on 7 reactors. SCORPIO was also installed on the Sizewell B reactor. Recently a new framework has been developed which further enhances the flexibility and capabilities for implementing core surveillance systems in different types of nuclear power plants. Modules can be added and replaced in an easy manner. It allows fast and reliable communication of data between modules based on the Software Bus tool developed by IFE. Further, the Picasso-3 user interface management system supports efficient implementation of different user interfaces. Both Unix and Windows NT platforms are supported. The new framework has been applied in development and installation of a SCORPIO-VVER version for the Dukovany NPP, Czech Republic. Here it was of particular importance to provide a flexible system for integration of modules originating from different companies. Development of a BWR version is now in progress. This means that SCORPIO will be available for all the major reactor types, and synergy is obtained by application of a common framework both with respect to system implementation and maintenance. By using the SCORPIO framework, the development time is reduced and the maintenance work is carried out more efficiently, compa