WorldWideScience

Sample records for core cooling ecc

  1. 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.)

  2. A study of ECC water bypass reduction technology for an improvement of core cooling capability

    International Nuclear Information System (INIS)

    Song, C. G.; Kwon, T. S.; Yun, B. J.

    2006-02-01

    The research for the reduction of ECC water bypass fraction was mainly performed to develop the flow mechanisms that ECC water can penetrate more effectively into a lower downcomer. Evaluation were carried out for the effect of major parameters on the bypass of ECC water in the downcomer with DVI. The following various physical models were derived for the reduction of the bypass fraction of ECC water: models of changing DVI injection angle, models of rearranging relative angles of DVI nozzles, model of grooved-and-subchannel type core barrel, model of dual core barrel. CFD analysis and MARS design verification were performed for the derived models as a first step performance estimation. Through out air-water verification experiments, quantitative evaluation were performed for each model, and three most efficient models were suggested. Examination were carried out for the requirement of structural modification and the change in structural integrity due to the adoption of one of the schemes

  3. 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

  4. 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

  5. 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.

  6. 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.)

  7. Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

    International Nuclear Information System (INIS)

    Hanson, G.H.

    1978-01-01

    The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. LTR 113-47 has shown that the LOFT ECCS can be safely bypassed or disabled when the total core power does not exceed 25 kW. A modified policy involves disabling the automatic actuation of the LOFT ECCS, but still retaining the manual activation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 70 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS

  8. Core-power and decay-time limits for disabled automatic-actuation of LOFT ECCS

    International Nuclear Information System (INIS)

    Hanson, G.H.

    1978-01-01

    The Emergency Core Cooling System (ECCS) for the LOFT reactor may need to be disabled for modifications or repairs of hardware or instrumentation or for component testing during periods when the reactor system is hot and pressurized, or it may be desirable to enable the ECCS to be disabled without the necessity of cooling down and depressurizing the reactor. A policy involves disabling the automatic-actuation of the LOFT ECCS, but still retaining the manual actuation capability. Disabling of the automatic actuation can be safely utilized, without subjecting the fuel cladding to unacceptable temperatures, when the LOFT power decays to 33 kW; this power level permits a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS. For the operating power of the L2-2 Experiment, the required decay-periods (with operating periods of 40 and 2000 hours) are about 21 and 389 hours, respectively. With operating periods of 40 and 2000 hours at Core-I full power, the required decay-periods are about 42 and 973 hours, respectively. After these decay periods the automatic actuation of the LOFT ECCS can be disabled assuming a maximum delay of 20 minutes following a LOCA for the manual actuation of ECCS. The automatic and manual lineup of the ECCS may be waived if decay power is less than 11 kW

  9. 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.)

  10. 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.)

  11. 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.)

  12. 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)

  13. Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

    International Nuclear Information System (INIS)

    Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

    2007-01-01

    Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

  14. 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.)

  15. 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

  16. Conclusions Drawn from the Investigation of LOCA-Induced Insulation Debris Generation and its Impact on Emergency Core Cooling (ECC) at German NPPs - Approach Taken by / Perspective of The German TSO (TuV)

    International Nuclear Information System (INIS)

    Huber, J.

    2004-01-01

    Initiated by the Barsebaeck incident in 1992 and the following activities related to the LOCA-induced insulation debris generation and its impact on emergency core cooling, investigations on German PWRs and BWRs were performed in these areas. The investigations on the German BWRs were carried out in detail immediately after the Barsebaeck incident in the years 1992 through 1994. Detailed investigations on the German PWRs started after the issue of the OECD report in 1996. Therefore the investigations on the impact of LOCA-induced insulation debris generation on strainer plugging carried out in Germany in the last years were focused mainly on the German PWRs. In the framework of these investigations of the German PWRs, the relevant parameters and phenomena were investigated in detail by the plant owners in the years from 1997 through 1999. The results were summarised in reports for each plant. The main results of the investigations conducted by the plant owners were that the plant owners considered backfitting in German PWRs is not necessary to guarantee emergency core cooling following a LOCA with insulation debris generation. As the technical support organisation for the German Bavarian and Hessian state authority, the TUV Suddeutschland was called upon to examine these investigations and the conclusions drawn by the plant owners. We compared each of the parameters and phenomena against the state of knowledge. The results of our examination in 1999 showed that the investigations of the plant owners were generally correct, but we stated also, that due to existing uncertainties, further investigations are necessary to validate the results. To meet these demands, the plant owners installed a working group for planning and performing newer, more realistic large-scale experiments (scaling factor 1:4) to investigate the transport mechanism of the insulation material within the containment sump, the head loss at the strainers and to estimate the amount of insulation

  17. 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.)

  18. 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

  19. 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.

  20. 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)

  1. 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.)

  2. 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.)

  3. Diablo Canyon ECCS enhancements

    International Nuclear Information System (INIS)

    Lin, A.; Lee, T.P.; Walter, L.E.

    2004-01-01

    Diablo Canyon Power Plant (DCPP) operated by Pacific Gas and Electric Co. (PG and E) is a Westinghouse designed four loop plant. In recent years, several issues were identified regarding the compliance of the Emergency Core Cooling System (ECCS) surveillance tests to the ECCS analyses assumptions. These concerns are related mostly to the High Head Safety Injection (HHSI) and the Intermediate Head Safety Injection (IHSI) systems where the injection line throttle valves are adjusted during outage surveillance testing to ensure compliance with the Technical Specifications (TS). To resolve all of the identified issues PG and E performed an ECCS reanalysis and upgraded the ECCS surveillance test program and also had Westinghouse perform a containment reanalysis using their latest model. As a result of these plant specific enhancement efforts, DCPP widened the operating window for TS surveillance testing, lowered the ECCS pumps' acceptance performance curves, and re-gained Peak Clad Temperature (PCT) and containment peak pressure margins. These enhancements are generically applicable to other plants and are addressed in this paper. (author)

  4. 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.)

  5. 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.)

  6. 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.)

  7. 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

  8. 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.)

  9. 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.)

  10. 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)

  11. 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

  12. Condensation during gravity driven ECC: Experiments with PACTEL

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

    This paper provides the results of the second series of gravity driven emergency core cooling (ECC) experiments with PACTEL (Parallel Channel Test Loop). The simulated accident was a small break loss-of-coolant accident (SBLOCA) with a break in a cold leg. The ECC flow was provided from a core makeup tank (CMT) located at a higher elevation than the main part of the primary system. The CMT was pressurized with pipings from the pressurizer and a cold leg. The tests indicated that steam condensation in the CMT can prevent ECC and lead to core uncovery.

  13. 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.

  14. 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)

  15. Emergency response guide-B ECCS guideline evaluation analyses for N reactor

    International Nuclear Information System (INIS)

    Chapman, J.C.; Callow, R.A.

    1989-07-01

    INEL conducted two ECCS analyses for Westinghouse Hanford. Both analyses will assist in the evaluation of proposed changes to the N Reactor Emergency Response Guide-B (ERG-B) Emergency Core System (ECCS) guideline. The analyses were a sensitivity study for reduced-ECCS flow rates and a mechanistically determined confinement steam source for a delayed-ECCS LOCA sequence. The reduced-ECCS sensitivity study established the maximum allowable reduction in ECCS flow as a function of time after core refill for a large break loss-of-coolant accident (LOCA) sequence in the N Reactor. The maximum allowable ECCS flow reduction is defined as the maximum flow reduction for which ECCS continues to provide adequate core cooling. The delayed-ECCS analysis established the liquid and steam break flows and enthalpies during the reflood of a hot core following a delayed ECCS injection LOCA sequence. A simulation of a large, hot leg manifold break with a seven-minute ECCS injection delay was used as a representative LOCA sequence. Both analyses were perform using the RELAP5/MOD2.5 transient computer code. 13 refs., 17 figs., 3 tabs

  16. Evaluation of mispositioned ECCS valves

    International Nuclear Information System (INIS)

    Hill, R.A.; O'Brien, J.F.; McIntire, D.C.; Barlow, R.T.

    1977-09-01

    In October of 1975, Westinghouse submitted NS-CE-787, dated October 17, 1975, to the Nuclear Regulatory Commission (NRC) and entered into discussions with them concerning the spurious movement of certain motor-operated valves (MOV's) in the Emergency Core Cooling System (ECCS) to a position defeating the ECCS function at a time when this function is required. On November 25, 1975, the discussion turned to the possible movement of a manually controlled, motor-operated valve due to a fault in its electrical circuitry and the NRC staff expressed concerns about other possible failure modes that might lead to such a valve movement. The NRC meeting minutes document these concerns. This report is an item-by-item response to the concerns expressed by the NRC staff at that meeting and incorporates the original electrical fault analysis

  17. 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)

  18. 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)

  19. 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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  1. 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

  2. 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

  3. Method and device for controlling ECCS

    International Nuclear Information System (INIS)

    Ikeda, Takashi; Kataoka, Yoshiyuki; Murase, Michio.

    1986-01-01

    Purpose: To accomplish reactor cooling without exposing fuel assemblies out of the coolant and also without inducing counter-current flow (CCFL) brake likely to be caused by excess injection of water even in case of malfunction of one system in a loss-of-coolant accident. Method: In a BWR type reactor having more than two independent ECCS, the lower plenum water level is measured and when the lower plenum is full of water, the ECCS are kept in a fully closed state, and reversely when the lower plenum is not full of water, more coolant than the lost quantity of water will be injected into the plenum at a higher pressure than a pressure at which fuel rods just begin to be exposed to the steam phase. The subcool energy of the emergency coolant to be injected is determined by the decay heat of the core and the change rate of a container pressure. However, the quantity of the emergency coolant is controlled such that the subcool energy will always become less than the overheating energy in the core range and the lower plenum range, thus improving safety and enabling the removal of a prior-art ECCS. (Kamimura, M.)

  4. 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.)

  5. 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)

  6. 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

  7. 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)

  8. ECCS Operability With One or More Subsystem(s) Inoperable

    International Nuclear Information System (INIS)

    Swantner, Stephen R.; Andrachek, James D.

    2002-01-01

    equivalent to a single Operable ECCS train exists with those components out of service. This evaluation ensures that the safety analysis assumption associated with one train of emergency core cooling system (ECCS) is still preserved by various combinations of components in opposite trains. An ECCS train is inoperable if it is not capable of delivering design flow to the reactor coolant system (RCS). Individual components are inoperable of they are not capable of performing their design function, or support systems are not available. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of Condition A is to maintain a combination of components such that 100% of the ECCS flow equivalent to a single Operable ECCS train remains available. This allows increased flexibility in plant operations under circumstances when components in the required subsystem may be inoperable, but the ECCS remains capable of delivering 100% of the required flow equivalent. This paper presents a methodology for identifying the minimum set of components necessary for 100% of the ECCS flow equivalent to a single Operable ECCS train. An example of the implementation of this methodology is provided for a typical Westinghouse 3-loop ECCS design. (authors)

  9. 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)

  10. Passive device for emergency core cooling of pressurized water reactors. Pasivno ustrojstvo za bezopasnost na vodo-voden atomen reaktor

    Energy Technology Data Exchange (ETDEWEB)

    Sikora, D

    1984-02-28

    The device proposed ensures additional margin of reactor subcriticality in case of post-accident emergency core cooling (ECC), using concentrated solution of chemical absorber and hot water from the secondary circuit. It consists of: a) a differential cylinder with a differential piston in it, with a lid and a seal, connected to a pipeline for secondary coolant; b) a pipeline for the secondary coolant; c) a volume between the lid and the piston for the secondary coolant from the steam generator; d) a discharge pipeline with a check valve of seal type connecting the inner volume of the differential cylinder to the discharge line; and e) a pipeline from the high-pressure volume of the differential cylinder filled with concentrated chemical absorber solution, to one of the main circulation loops. The device permits ECC innovation of the operating non-standard nuclear power plants with PWR type reactors.

  11. 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.

  12. Effects of high temperature ECC injection on small and large break BWR LOCA simulation tests in ROSA-III program (RUNs 940 and 941)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Kumamaru, Hiroshige; Anoda, Yoshinari; Yonomoto, Taisuke; Murata, Hideo; Tasaka, Kanji

    1990-03-01

    The ROSA-III program, of which principal results are summarized in a report of JAERI 1307, conducted small and large-break loss-of-coolant experiments (RUNs 940 and 941) with high water temperature of the emergency core cooling system (ECCS) are one of the parametric study with respect to the ECCS effect on core cooling. This report presents all the experiment results of these two tests and describes additional finding with respect to the hot ECC effects on core cooling phenomena. By comparing these two tests (water temperature of 393 K) with the standard ECC tests of RUNs 922 and 926 (water temperature of 313 K), it was found that the ECC subcooling variation had a small influence on the core cooling phenomena in 5 % small break tests but had larger influence on them in 200 % break tests. The ECC subcooling effects described in the previous report are reviewed and the temperature distribution in the pressure vessel is investigated for these four tests. (author)

  13. The Heat Exchanger for Passive Part ECCS of WWER-1000 on Base of the Thermo siphons

    International Nuclear Information System (INIS)

    Kirov, V.; Chulkin, O.

    2008-01-01

    One of NPP's systems providing safe operation is the system of emergency core cooling system (ECCS), which primary function in accidents is to flood the nuclear reactor core and to assure the sub critical condition and core cooling. At injection of cold water in reactor thermal stresses and thermal fatigue in the vessel cladding and constructional materials are arise. Low temperature of the water injected in reactor is a reason of occurrence of these undesirable consequences. Some variants of the water heating in accumulators of ECCS are considered. Now at Ukrainian NPPs the electrical heating in accumulators is used. Electrical heaters create the essential additional loading to diesel generators at imposing of two accidents - the large break and losses of power supplies on own needs. It is offered to use a heater in accumulators that working by a principle two-phase thermal siphon which advantages is: small dimensions, small delay and design reliability. In such heat exchanger the heating medium is a direct steam and the heated up medium is water with boric acid from accumulators of ECCS. Under requirements of the service regulations of ECCS accumulators it is necessary to guarantee injected water heating up to 90 ?? in case of a small break and to 150 ?? in case of the large break. Results of calculations for different external diameters of a tube of thermal siphon which have allowed to define the constructive sizes of heat exchanger, providing necessary conditions for required functioning of passive part ECCS are submitted The calculation and analysis of operating modes of the changed circuit of passive part ECCS for various accidents is carried out. The calculated pressure drop indicates that changes do not have essential influence on system work as a whole. Thus, the submitted decision provides the increase of reliability of ECCS at small and large breaks accidents, i.e. in all modes stipulated by the project.(author)

  14. Advanced DVI for ECC direct bypass mitigation

    International Nuclear Information System (INIS)

    Kwon, Tae-Soon; Song, Chul-Hwa; Baek, Won-Pil

    2009-01-01

    An ECC direct bypass fraction during a late reflood phase of a LBLOCA is strongly dependent on the characteristics of the cross flow and the geometrical configuration of a DVI in the downcomer of a pressurized light water reactor. The important design parameters of a DVI are the elevation, the azimuthal angle, and the separator to prevent a steam-water interaction. An ECC sub-channel to separate or to isolate an ECC water from a high-speed cross flow is one of the important design features to mitigate the ECC bypass phenomena. A dual core barrel cylinder as an ECC flow separator is located between a reactor vessel and a core barrel outer wall in the downcomer annulus. A new narrow gap between the core barrel and the additional dual core barrel plays the role of a downward ECC flow channel or an ECC flow separator in a high-speed cross flow field of the downcomer annulus. The flow zone around a broken cold leg in the downcomer annulus has the role of a high ECC direct bypass due to a strong suction force while the wake zone of a hot leg has the role of an ECC penetration. Thus, the relative azimuthal angle of the DVI nozzle from the broken cold leg is an important design parameter. A large azimuthal angle from a cold leg to a hot leg needs to avoid a high suction flow zone when an ECC water is being injected. The other enhancing mechanism of an ECC penetration is a grooved core barrel which has small rectangular-shaped grooves vertically arranged on the core barrel wall of the reactor vessel downcomer annulus. These grooves have the role for a generation of a vortex induced by a high-speed cross flow. Since the stagnant flow in a lateral direction and rotational vortex provides the pulling force of an ECC drop or film to flow down into the lower downcomer annulus by gravity, the ECC direct bypass fraction is reduced when compared to the current design of a smoothed wall. An open channel of grooves generates a stagnant vortex, while a closed channel of grooves

  15. Development of Calculation Algorithm for ECCS Kinematic Shock

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

    The void fraction of inverted U-pipes in front of SI(Safety Injection) pumps impact on the pipe system of ECCS(Emergency Core Cooling Systems). This phenomena is called as 'Kinematic Shock'. The purpose of this paper is to achieve the more exactly calculation when the kinematic shock is calculated by simplified equation. The behavior of the void packet of the ECCS pipes is illustrated by the simplified (other name is kinematic shock equation).. The kinematic shock is defined as the depth of total length of void clusters in the pipes of ECCS when the void cluster is continually reached along the part of pipes in vertical direction. In this paper, the simplified equation is evaluated by comparing calculation error each other.]. The more exact methods of calculating the depth of the kinematic shock in ECCS is achieved. The error of kinematic shock calculation is strongly depended on the calculation search gap and the order of Taylor's expansion. From this study, to select the suitable search gap and the suitable calculation order, differential root method, secant method, and Taylor's expansion form are compared one another.

  16. 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

  17. Safety evaluation report on Westinghouse Electric Company ECCS evaluation model for plants equipped with upper head injection

    International Nuclear Information System (INIS)

    Lauben, G.N.; Wagner, N.H.; Israel, S.L.; McPherson, G.D.; Hodges, M.W.

    1978-04-01

    For plants which include an ice condenser containment concept, Westinghouse has planned an additional safety system known as the upper head injection (UHI) system to augment the emergency core cooling system. This system is comprised of additional accumulator tanks and piping arranged to supply cooling water to the top of the core during the blowdown period following a postulated large-break loss-of-coolant accident (LOCA). The objective of UHI is to add to the core cooling provided by the conventional emergency core cooling system (ECCS) and so permit operation at linear heat rates comparable to those permitted in plants utilizing the dry containment concept. In this way, plants which include the UHI system would have greater operating flexibility while still meeting the acceptance criteria as defined in paragraph 50.46 of 10 CFR Part 50. This review is concerned with those changes to the Westinghouse ECCS evaluation model that have been proposed for the UHI-LOCA model

  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. Historical information on ORNL proposals for ECCS testing

    International Nuclear Information System (INIS)

    1976-01-01

    This document contains a compilation of the correspondence and preliminary report which the Oak Ridge National Laboratory submitted to the U.S. Atomic Energy Commission on the subject of in situ testing of emergency core cooling systems (ECCS) in pressurized water reactors. Most of the correspondence, which was prepared four to five years ago, deals with an ORNL proposal to evaluate the merits of conducting an in situ ECCS test in a full-scale commercial power plant. The end result of this work was to be a report in which the feasibility of conducting such an experiment would be discussed and a ''rough'' cost estimate provided. In support of this proposal, ORNL prepared a brief preliminary prospectus which identified some of the key questions that were to be addressed should the evaluation study be funded

  20. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  1. Accomplishments of LOCA/ECCS experimental research at Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Murao, Yoshio; Koizumi, Yasuo

    1984-01-01

    Japan Atomic Energy Research Institute has investigated loss-of-coolant accident (LOCA)/emergency core cooling system (ECCS) from 1970. Major results of the LOCA/ECCS research are summarized in this report. ROSA-II program was LOCA/ECCS research for a pressurized water reactor (PWR) and ROSA-III program was for a boiling water reactor (BWR). The both test facilities were scaled at approximately 1/400 of the respective reference PWR and BWR. Large scale reflood test is research on reflood phenomena during a large break LOCA of PWR. The test facility is scaled at approximately 1/20 of the reference PWR and the research is still being continued. (author)

  2. 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)

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. Investigation of Characteristics of Impinging Jet for 1/5-Scale ECC injection

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Byung Soo; Ko, Yung Joo; Bae, Hwang; Kwon, Tae Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    In ECCS of SMART reactor, safety injection pump discharges cooling water into the core to maintain the water level by filling the amount of loss of coolant under emergency situation such as SBLOCA. Once the ECCS starts to operate, the injected water will be impinged to the upper wall of core support barrel (CBS). And the water will fall along the wall forming liquid film or droplets as shown in Fig. 1(b) due to high Reynolds number. The breakup and film flow will be bypassed by high temperature and pressure steam-water mixture cross flow from RCP discharge into the atmosphere through broken injection nozzle. Then, the flow phenomena in the downcomer is very complex situation with including jet impingement, jet breakup, liquid entrainment, steam condensation, counter-current flow and etc. In this study, the hydraulic features of impinging jet were investigated through visualization for full scale test for simulation of SMART ECC jet and SWAT test of 1/5 simulated test for ECCS of SMART reactor and measurement of the film width. And the scaling method for SWAT test was discussed considering jet break up and other phenomena

  8. A study on design of the trip computer for ECCS based on dynamic safety system

    International Nuclear Information System (INIS)

    Kim, Seog Nam

    2000-02-01

    The Emergency Core Cooling system in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relay and analog comparator logic which is difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System is implemented. The Dynamic Safety System (DSS) is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved between the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this thesis, a possible implementation of the DSS using PLC is presented for a CANDU reactor. ECC System of the CANDU Reactor is selected as the reference system. The function of the DSS is implemented In PLC with the CONCEPT language. CONCEPT was developed by GROUPE SCHNEIDER as a graphic user interface programming tool for the Quantum PLC. A MMI display for ECCS based on DSS is implemented with LOOKOUT as an object driven programming tool. The Validation test has been performed by S/W Input Simulator as per Validation Test Procedure. The result of the test was checked and displayed on the MMI display. From the test results, it is shown that the DSS based ECC System operates correctly in all conditions

  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. 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.

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

    Science.gov (United States)

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

    2016-05-01

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

  12. 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.)

  13. 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)

  14. Evaluation of ECC bypass data with a nonlinear constrained MLE technique

    International Nuclear Information System (INIS)

    Bishop, T.A.; Collier, R.P.; Kurth, R.E.

    1980-01-01

    Recently, Battelle's Columbus Laboratories have been involved in scale-model tests of emergency core cooling (ECC) systems for hypothesized loss-of-coolant accidents in pressurized water reactors (PWR). These tests are intended to increase our understanding of ECC bypass, which can occur when steam flow from the reactor core causes the emergency coolant to bypass the core and flow directly to the break. One objective of these experiments is the development of a correlation which relates the flow rate of water penetrating to the core to the steam flow rate. This correlation is derived from data obtained from a 2/15 scale model PWR at various ECC water injection rates, subcoolings, pressures, and steam flows. The general form of the correlation being studied is a modification of the correlation first proposed by Wallis. The correlation model is inherently nonlinear and implicit in form, and the model variables are all subject to error. Therefore, the usual nonlinear analysis techniques are inappropriate. A nonlinear constrained maximum-likelihood-estimation technique has been used to obtain estimates of the model parameters, and a Battelle-developed code, NLINMLE, has been used to analyze the data. The application of this technique is illustrated by sample calculations of estimates of the model parameters and their associated confidence intervals for selected experimental data sets. 5 figures, 7 tables

  15. 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

  16. 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...

  17. 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

  18. Minimum containment pressure and its effect on ECCS performance of APR-1400

    International Nuclear Information System (INIS)

    Kim, In Goo; Bang, Young S.; Kim, Hho Jung

    2004-01-01

    The containment pressure has a strong effect on the late reheat behavior for a large break LOCA, associated with the DVI issue. The downcomer boiling, which occurs during the post-reflood phase, has a negative effect on core cooling for a LBLOCA. Because the downcomer boiling is enhanced as the containment pressure decreases, how to determine containment pressure is important to the evaluation of ECCS performance. In spite of its importance of containment pressure, there are few studies on the containment pressure and the interaction between RCS and containment thermal hydraulics. To have a better knowledge of the effect of containment pressure on APR-1400 ECCS performance, a parametric study for containment pressure has been carried out. Also, the interaction between RCS and containment behavior has been also investigated

  19. 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)

  20. 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.)

  1. ECC cusp analysis

    Energy Technology Data Exchange (ETDEWEB)

    Pregliasco, R.G. (Centro Atomico Bariloche and CONICET, S.C. de Bariloche (Argentina)); Garibotti, C.R. (Centro Atomico Bariloche and CONICET, S.C. de Bariloche (Argentina)); Barrachina, R. (Centro Atomico Bariloche and CONICET, S.C. de Bariloche (Argentina))

    1994-03-01

    We measured the ionization double differential cross section for the H[sup +][yields]He 100 keV collision at the proximities of the ECC cusp and all around 0 . We propose an angular function expansion and a procedure for data analysis. We demonstrate that, in this way, the difficulties and shortcomings presented in the standard data processing are overcome. In our experimental results we distinguish the presence of characteristic shapes for the ionization depending on whether the module of the electron velocity is greater or smaller than the projectile velocity. (orig.)

  2. 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

  3. 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.

  4. 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.

  5. 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.

  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. 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.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  10. 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...

  11. 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.

  12. 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)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

    emergency core cooling (ECC) water entering the RPV through the ECC injection into the cold leg. The experimental results show an incomplete mixing with typical concentration and temperature distributions at the core inlet, which strongly depend on the boundary conditions. Computational fluid dynamics calculations were found to be in good agreement with the experiments

  14. 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.)

  15. 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)

  16. 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)

  17. 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.

  18. 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.

  19. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  20. 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.)

  1. 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

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

    International Nuclear Information System (INIS)

    Lakkis, I.A.

    1993-01-01

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

  3. 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.

  4. 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

  5. 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.)

  6. 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

  7. 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.

  8. Experiment data report for Semiscale Mod-1 Test S-05-3 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

    1977-03-01

    Recorded test data are presented for Test S-05-3 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-3 was conducted from initial conditions of 2263 psia and 545 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg sides of the intact and broken loops and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. For Test S-05-3, specifically the effects of upper plenum coolant injection on core thermal and system response were being investigated

  9. 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

  10. 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.

  11. 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...

  12. Safety verification for the ECCS driven by the electrically 4 trains during LBLOCA reflood phase using ATLAS

    International Nuclear Information System (INIS)

    Park, Yusun; Park, Hyun-sik; Kang, Kyoung-ho; Choi, Nam-hyun; Min, Kyoung-ho; Choi, Ki-yong

    2014-01-01

    Highlights: • Safety improvement by adopting 4 train emergency core cooling system was validated experimentally. • General thermal hydraulic behaviors of the system during LBLOCA reflood phase were successfully demonstrated. • Key parameters such as the liquid levels, the PCTs, the quenching time, and the ECC bypass ratios were investigated. • Asymmetric effects of the different combination of safety injection were negligible during the reflood period. - Abstract: The APR1400 is equipped with four safety injection pumps driven by two emergency diesel generators. However, the design has been changed so that the four safety injection pumps are driven by 4 emergency diesel generators during the design certification process from the U.S. NRC. Thus, 4 safety injection pumps (SIPs) are completely independent electrically and mechanically and three safety injection pumps are available in a single failure condition. This design change could have a certain effects on the thermal-hydraulic phenomenon occurring in the downcomer region during the late reflood phase of a large break loss of coolant accident (LBLOCA). Thus, in this study, a verification experiment for the reflood phase of a LBLOCA was performed to evaluate the core cooling performance of the 4 train emergency core cooling system (ECCS) with an assumption of a single failure. And the different combinations of three SIPs positions were tested to investigate the asymmetric effects on the reactor core cooling performance. The overall experimental results revealed the typical thermal–hydraulic trends expected to occur during the reflood phase of a large-break LOCA scenario for the APR1400. Experiment with the injection of three SIPs showed a faster core quenching time and lower bypass ratio than that of the case in which two SIPs were injected. The RPV wall temperature distributions showed the similar trend in spite of the different SIP combinations

  13. 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)

  14. 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)

  15. 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

  16. 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)

  17. Experiment data report for semiscale Mod-1 test S-04-1 (baseline ECC test)

    International Nuclear Information System (INIS)

    Crapo, H.S.; Collins, B.L.; Sackett, K.E.

    1976-09-01

    Recorded test data are presented for Test S-04-1 of the Semiscale Mod-1 Baseline ECC Test Series. This test is among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-1 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using system volume scaled coolant injection parameters. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW in such a manner as to simulate the surface heat flux response of nuclear fuel rods until such time that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core cooling injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown

  18. 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

  19. 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

  20. 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)

  1. 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

  2. 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

  3. Parametric Effects of Debris Source, Environments, and Design Options on the Overall Performance of ECCS Recirculation Sump

    International Nuclear Information System (INIS)

    Park, Jong Woon; Kim, Chang Hyun

    2006-01-01

    A primary safety issue regarding long-term recirculation core cooling following a LOCA (Loss of Coolant Accident) is that LOCA-generated debris may be transported to the recirculation sump screen, resulting in adverse blockage on the sump screen and deterioration of available NPSH (Net Positive Suction Head) of ECCS (Emergency Core Cooling System). USNRC identified this as Generic Safety Issue (GSI) 191 and issued the Generic Letter 04-02 to resolve the issue. The GL required that all PWR owners perform an engineering assessment of their containment recirculation sumps to ensure they will not suffer from excessive blockage. The guidance report (GR) for PWR sump performance evaluation has been developed by NEI (Nuclear Energy Institute) and approved by the USNRC. In Korea, Korea Hydro and Nuclear Power Company (KHNP) is performing the assessment of Kori unit 1 and planning for remaining plants in the near future. The objective of the assessment is to derive required plant modifications including insulation, sump screen, etc. To derive the cost-effective modification items, we have to get insight on the parametric effects of plant conditions and design. Therefore, the general effects of debris source, containment environments and debris interceptor on the performance of ECCS recirculation sump with respect to head loss are parametrically investigated

  4. 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.

  5. Design Development of SMART ECC Water Asymmetric Two-phase choking test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Il; Cho, Seok; Ko, Yung Joo; Shin, Yong Cheol; Kwon, Tae Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    SMART pressurized water reactor type is different from the existing integral NSSS commercial pressurized water reactor system which is equipped with the main features. In addition RCS piping is removed and the feature of the SBLOCA is a major design break accident. The TASS / SMR code is analyzed SMART SBLOCA. In order to verify analysis code, SMART analysis for verification of conservatism is promoting using data for experiments with Integral Effect Test and Separate Effect. In this paper, the design feature of the SWAT (SMART ECC Water Asymmetric Two-phase choking test facility) is described. SWAT is linearly reduced to a 1/5 ratio while the geometrical shape is conserved. In major shape of SMART ECC injection performance test, distortions which caused by gravitational effects are minimized. Because both the emergency core cooling water injection nozzle height and the break nozzle height match the RCP Suction Nozzle height in test section of the main forms. The main part of the test section is SG-side upper down-comer. The boundary conditions are saturated steam and water flow condition and drain flow rate to control the collapsed water level in the down-comer

  6. Large break LOCA analysis for retrofitted ECCS at MAPS using modified computer code ATMIKA

    International Nuclear Information System (INIS)

    Singhal, Mukesh; Khan, T.A.; Yadav, S.K.; Pramod, P.; Rammohan, H.P.; Bajaj, S.S.

    2002-01-01

    Full text: Computer code ATMIKA which has been used for thermal hydraulic analysis is based on unequal velocity equal temperature (UVET) model. Thermal hydraulic transient was predicted using three conservation equations and drift flux model. The modified drift flux model is now able to predict counter current flow and the relative velocity in vertical channel more accurately. Apart from this, stratification model is also introduced to predict the fuel behaviour under stratified condition. Many more improvements were carried out with respect to solution of conservation equation, heat transfer package and frictional pressure drop model. All these modifications have been well validated with published data on RD-12/RD-14 experiments. This paper describes the code modifications and also deals with the application of the code for the large break LOCA analysis for retrofitted emergency core cooling system (ECCS) being implemented at Madras Atomic Power Station (MAPS). This paper also brings out the effect of accumulator on stratification and fuel behaviour

  7. Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

    1977-02-01

    Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-1 was conducted from initial conditions of 2263 psia and 544 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

  8. Experiment data report for Semiscale Mod-1 Test S-05-5 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

    1977-04-01

    Recorded test data are presented for Test S-05-5 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-5 was conducted from initial conditions of 2263 psia and 537 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. The upper plenum was vented through a reflood bypass line interconnecting the hot and cold legs of the broken loop

  9. 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

  10. Method of controlling ECCS system in reactors

    International Nuclear Information System (INIS)

    Oohashi, Hideaki; Ikehara, Morihiko.

    1982-01-01

    Purpose: To eliminate the risk of misoperation and thereby improve the reliability of ECCS system upon accident. Method: ECCS system for nuclear reactor is automatically started by either of signals from a water level detector in a pressure vessel or from a pressure detector in a reactor container. Further, the ECCS system is started or stopped by the manual operation irrespective of the signals, and the signals from the pressure detector are isolated from the ECCS-starting signal by the contacts which actuate interlocked with the stopping operation of the manual operation switch. Then, after stopping the ECCS system by the manual operation, the ECCS system is started by the signals from the water level detector irrespective of the signals from the pressure detector. (Seki, T.)

  11. 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)

  12. Evaluation report on SCTF Core-III Test S3-22

    International Nuclear Information System (INIS)

    Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; Akimoto, Hajime; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

    1991-07-01

    Two tests (Tests S3-20 and S3-22) were conducted with JAERI's Slab Core Test Facility (SCTF) Core-III in order to investigate water break-through and core cooling behaviors under the intermittent ECC water delivery from the hot legs to one location in the upper plenum and the alternate ECC water delivery to two locations in the upper plenum during reflooding, respectively. This report presents an analysis on Test S3-22 (the alternate case). Subcooled ECC water was injected alternately just above the upper core support plate above Bundles 7 and 8 and Bundles 3 and 4. The total injection rate from both injection ports was the same as that in SCTF Test S3-20 and Test S3-13. Analyzing the test data together with those of Tests S3-13 and S3-20 the following has been found: (1) Alternate break-through occurred immediately corresponding to the alternate ECC water injection except for one period, during which no break-through was observed. However, there observed a difference in break-through behavior that break-through was strong above the low power region, whereas weak above the high power region. (2) Although its break-through behavior was different, nearly the same core cooling as in the continuous or intermittent ECC water delivery case was observed except for the period around quench. (3) Around quench time, degraded core cooling comparing to the continuous or intermittent ECC water delivery case was observed. That is, quench time at the midplane level of the present test was 35 s later than in the continuous case. This is considered to result from decrease in core water inventory caused by water sealing at the cross-over leg. (J.P.N.)

  13. 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.

  14. 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

  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. 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)

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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)

  4. Performance-based ECCS cladding acceptance criteria: A new simulation approach

    International Nuclear Information System (INIS)

    Zoino, A.; Alfonsi, A.; Rabiti, C.; Szilard, R.H.; Giannetti, F.; Caruso, G.

    2017-01-01

    Highlights: • A new methodology to demonstrate compliance with the new ECCS acceptance criteria is described. • A wide spectrum of fuel rod initial burnup states can be analysed in the design phase. • The coupled suite PHISICS/RELAP5-3D has been used in the analyses. • A demo simulation of the equilibrium cycle, load-following and a LOCA analysis has been performed. - Abstract: The U.S. Nuclear Regulatory Commission is currently proposing rulemaking to revise the Loss Of Coolant Accident (LOCA) and therefore the 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. 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 study here presented has been part of a big project used to investigate technical issues and approaches for future industrial applications within the Risk-Informed Safety Margin Characterization (RISMC) Pathway. Specifically, the primary aim of this study is to lay out a roadmap to demonstrate the application of the new methodology. The present analysis shows a simplified version of the methodology of an industrial application on the Core Design and the Multi-Cycle Analysis.

  5. A study on transient heat transfer of the EU-ABWR external core catcher using the phase-change effective convectivity model

    International Nuclear Information System (INIS)

    Tran Chi Thanh; Nguyen Viet Hung; Tahara, Mika; Kojima, Yoshihiro; Hamazaki, Ryoichi; Kudinov, Pavel

    2015-01-01

    In advanced designs of Nuclear Power Plants (NPPs), for mitigation of severe accident consequences, on the one hand, the In-Vessel Retention (IVR) concept has been implemented. On the other hand in other new NPP designs (Generation III and III+) with large power reactors, the External Core Catcher (ECC) has been widely adopted. Assessment of ECC design robustness is largely based on analysis of heat transfer of a melt pool formed in the ECC. Transient heat transfer analysis of an ECC is challenging due to (i) uncertainty in the in-vessel accident progression and subsequent vessel failure modes; (ii) long transient, (iii) high Rayleigh number and complex flows involving phase change of the melt pool formed in an ECC. The present paper is concerned with analysis of transient melt pool heat transfer in the ECC of new Advanced Boiling Water Reactor (ABWR) designed by Toshiba Corporation (Japan). According to the ABWR severe accident management strategy, the ECC is initially dry. In order to prevent steam explosion flooding is initiated after termination of melt relocation from the vessel. The ECC full of melt is cooled from the top directly by water and from the bottom through the ECC walls. In order to assess sustainability of the ECC, heat transfer simulation of a stratified melt pool formed in the ECC is carried out. The problem addressed in this work is heat flux distribution at ECC boundaries when cooling is applied (i) from the bottom, (ii) from the top and from the bottom. To perform melt pool heat transfer simulation, we employ Phase-change Effective Convectivity Model (PECM) which was originally developed as a computationally efficient, sufficiently accurate, 2D/3D accident analysis tools for simulation of transient melt pool heat transfer in the reactor lower plenum. Thermal loads from the melt pool to ECC boundaries are determined for selected ex-vessel accident scenarios. Performance of the ECC, efficiency of severe accident management (SAM) measures and

  6. 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

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

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

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

  8. 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

  9. 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)

  10. Experiments on injection performance of SMART ECC facility using SWAT

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Il; Cho, Seok; Ko, Yung Joo; Min, Kyoung Ho; Shin, Yong Cheol; Kwon, Tae Soon; Yi, Sung Jae; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    SMART (System-integrated Modular Advanced ReacTor), an advanced integrated PWR is now in the under developing stages by KAERI. Such integral PWR excludes large-size piping of the primary system of conventional PWR and incorporates the SGs into RPV, which means no LBLOCA could occur in SMART. Therefore, the SBLOCA is considered as a major DBA (Design Basis Accident) in SMART and it is mainly analyzed by using TASS/SMR computer code. The TASS/SMR code should be validated using experimental data from both Integral Effect Test and Separate Effect Test facilities. To investigate injection performance of the ECC system, on SET facility, named as SWAT (SMART ECC Water Asymmetric Two-phase choking test facility), has been constructed at KAERI. The SWAT simulates the geometric configurations of the SG-side upper downcomer annulus and ECCSs of those of SMART. It is designed based on the modified linear scaling method with a scaling ratio of 1/5, to preserve the geometrical similarity and minimize gravitational distortion. The purpose of the SWAT tests is to investigate the safety injection performance, such as the ECC bypass in the downcomer and the penetration rate in the core during the SBLOCA, and hence to produce experimental data to validate and the prediction capability of safety analysis codes, TASS/SMR

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. A PWR reactor downcomer modification for reduction of ECC bypass flow during LOCA

    International Nuclear Information System (INIS)

    Popov, N.; Bosevski, T.

    1986-01-01

    The ECC bypass phenomenon in the PWR reactor down-comer, which delays the reactor vessel refilling, after cold leg large break LOCA accident, has been subject of analysis in this paper. In the paper, a particular construction modification of the reactor down-comer has been suggested by inserting vertical ribs, aimed to intensify the reactor ECC refilling following the LOCA accident, and to advance the thermal-hydraulics safety of post-accidental cooling of the PWR reactors. To verify the effectiveness of the suggested down-comer construction modification, some properly selected results, obtained by corresponding verified mathematical model, have been presented in this paper. (author)

  2. 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.

  3. Evaluation report on SCTF Core-III test S3-06

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Ohnuki, Akira; Adachi, Hiromichi; Murao, Yoshio; Minato, Akihiko; Sakaki, Isao.

    1988-10-01

    In order to investigate the effect of radial power distribution on the thermal-hydraulic characteristics during the reflood phase of a PWR-LOCA with a combined injection type ECCS, a core cooling separate effect test S3-06 and a combined injection test S3-16-Phase 2 were performed using the Slab Core Test Facility (SCTF) Core-III. The radial power distributions in these two tests simulated a reference distribution for a PWR with a combined injection type ECCS and a steep distribution for a PWR with a cold leg injection type ECCS, respectively. Under the radial power distribution of a PWR with a combined injection type ECCS, the radial power distribution had little effect on the thermal-hydraulic behavior in the two-phase up-flow region due to the approximately flat power distribution in this region (power ratio = 1.04 ∼ 1.08). The overall fluid behavior in the pressure vessel was also little affected by the radial power distribution. On the other hand, under the steep radial power distribution (peak power ratio = 1.36), the degree of heat transfer enhancement in high power bundles in the two-phase up-flow region was dominated by the bundlewise radial power ratio as in the case of a PWR with a cold leg injection type ECCS. (author)

  4. 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.

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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)

  12. 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.

  13. 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

  14. 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...

  15. 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

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. 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

  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. 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

  3. ECCS analysis of B and W's 177-FA lowered-loop NSS

    International Nuclear Information System (INIS)

    Jones, R.C.; Biller, J.R.; Dunn, B.M.

    1975-06-01

    The effectiveness for the ECCS for B and W's 177-FA Lowered-Loop NSS is shown to meet the five criteria of 10 CFR 50.46. A spectrum analysis is performed and the worst case is used to establish allowable linear heat rates as a function of elevation in the core. (U.S.)

  4. 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

  5. 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)

  6. 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)

  7. Multi-dimensional analysis of the ECC behavior in the UPI plant Kori Unit 1

    International Nuclear Information System (INIS)

    Bae, Sungwon; Chung, Bub-Dong; Bang, Young Seok

    2008-01-01

    A multi-dimensional transient analysis during the LBLOCA of the Kori Unit 1 has been performed by using the MARS code. Based on 1-D nodalization of the Kori Unit 1, the reactor vessel nodalizations have been replaced by the multi-dimensional component. The multi-dimensional component for the reactor vessel is designed as 5 radial, 8 peripheral, and 21 vertical grids. It is assumed that the fuel assemblies are homogeneously distributed in inner 3 radial grids. The outer 1 radial grid region is modeled as the core bypass. The outer-model 1 radial grid is used for the downcomer region. The corresponding heat structures and fuels are modified to fit for the multi-dimensional reactor vessel model. The form drag coefficients for the upper plenum and the core have been designated as 0.6 and 9.39, respectively. The form drag coefficients for the radial and peripheral directions are assigned to the same on the assumption of homogeneous distribution of the flow obstacles. After obtaining the 102% power steady operation condition, cold leg LOCA simulation is performed during 400 second period. The multi-dimensional steady run results show no severe differences compared to the traditional 1-D nodalization results. After the ECC injection starts, a liquid pool is maintained at the upper plenum because the ECCS water can not overcome the upward gas flow that comes from the reactor core through the upper tie plate. The depth of ECCS water pool is predicted as about 20% of the total height from the upper tie plate and the center line of the hot leg pipe. At the vicinity region of the active ECCS show higher depth of liquid pool. The accumulated water flow rate passing the upper tie plate is calculated by the transient result. Much downward water flow is obtained at the outer-most region of upper plenum space. The downward flow dominant region is about 32.3% of the total upper tie plate area. The accumulated ECCS bypass ratio is predicted as 27.64% at 300 second. It is calculated

  8. 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)

  9. ECC2K-130 on NVIDIA GPUs

    NARCIS (Netherlands)

    Bernstein, D.J.; Chen, H.-C.; Cheng, C.M.; Lange, T.; Niederhagen, R.F.; Schwabe, P.; Yang, B.Y.

    2012-01-01

    [Updated version of paper at Indocrypt 2010] A major cryptanalytic computation is currently underway on multiple platforms, including standard CPUs, FPGAs, PlayStations and GPUs, to break the Certicom ECC2K-130 challenge. This challenge is to compute an elliptic-curve discrete logarithm on a Koblitz

  10. ECC2K-130 on NVIDIA GPUs

    NARCIS (Netherlands)

    Bernstein, D.J.; Chen, H.-C.; Cheng, C.M.; Lange, T.; Niederhagen, R.F.; Schwabe, P.; Yang, B.Y.; Gong, G.; Gupta, K.C.

    2010-01-01

    A major cryptanalytic computation is currently underway on multiple platforms, including standard CPUs, FPGAs, PlayStations and Graphics Processing Units (GPUs), to break the Certicom ECC2K-130 challenge. This challenge is to compute an elliptic-curve discrete logarithm on a Koblitz curve over $\\rm

  11. 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.

  12. 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)

  13. 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)

  14. Evaluation report on SCTF Core-III tests S3-7 and S3-8

    International Nuclear Information System (INIS)

    Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi

    1990-03-01

    It has been said that the Emergency Core Cooling (ECC) water injected into the hot legs flows into the upper plenum and then falls back to the core (i.e. break-through) during reflood phase in a German type Pressurized Water Reactor (GPWR) with the combined-injection-type ECCS, and that the break-through occurs where the water temperature at the tie plate area is lower and subcooled. Based on this information two tests were conducted with the Slab Core Test Facility (SCTF) Core-III in order to investigate the effects of the water temperature distribution at the tie plate area on the break-through and the core cooling. In these tests, the subcooled ECC water was injected just above the Upper Core Support Plate (UCSP) in order to establish the desired water temperature distribution at the tie plate area. In one test (Test S3-7) the ECC water injection above the UCSP was performed above Bundles 3 and 4, and in the other test (Test S3-8) above Bundles 7 and 8 during initial 60 s a and then was changed to above Bundles 3 and 4. The test data were compared with those of Test S3-SH1, in which the injection was performed above Bundles 7 and 8 and the other test conditions were the same as in Tests S3-7 and S3-8. Analyzing these test data, the following has been found: The break-through occurs where the water temperature at the tie plate area is subcooled and the core cooling is enhanced significantly in the break-through region. The break-through location changes, with some time lag, following the change of the water temperature distribution at the tie plate area. Furthermore, the core cooling in the non-break-through regions is almost the same regardless of the location of the break-through. (author)

  15. Direct vessel inclined injection system for reduction of emergency core coolant direct bypass in advanced reactors

    International Nuclear Information System (INIS)

    Yoon, Sang H.; Lee, Jong G.; Suh, Kune Y.

    2006-01-01

    Multidimensional thermal hydraulics in the APR1400 (Advanced Power Reactor 1400 MWe) downcomer during a large-break loss-of-coolant accident (LBLOCA) plays a pivotal role in determining the capability of the safety injection system. APR1400 adopts the direct vessel injection (DVI) method for more effective core penetration of the emergency core cooling (ECC) water than the cold leg injection (CLI) method in the OPR1000 (Optimized Power Reactor 1000 MWe). The DVI method turned out to be prone to occasionally lack in efficacious delivery of ECC to the reactor core during the reflood phase of a LBLOCA, however. This study intends to demonstrate a direct vessel inclined injection (DVII) method, one of various ideas with which to maximize the ECC core penetration and to minimize the direct bypass through the break during the reflood phase of a LBLOCA. The 1/7 scaled down THETA (Transient Hydrodynamics Engineering Test Apparatus) tests show that a vertical inclined nozzle angle of the DVII system increases the downward momentum of the injected ECC water by reducing the degree of impingement on the reactor downcomer, whereby lessening the extent of the direct bypass through the break. The proposed method may be combined with other innovative measures with which to ensure an enough thermal margin in the core during the course of a LBLOCA in APR1400

  16. 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.)

  17. 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)

  18. 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.

  19. 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

  20. Water-hammer in the cold leg during an SBLOCA due to cold ECCS injection

    International Nuclear Information System (INIS)

    Ortiz, M.G.; Ghan, L.S.

    1991-01-01

    Water-hammer might occur in the cold leg of pressurized water reactors (PWR) during small break loss-of-coolant accidents (SBLOCA's), when cold emergency core cooling system (ECCS) water is injected into a pipe that may be partially filled with saturated steam. The water may mix with the steam and cause it to condense abruptly. Depending on the flow regime present, slugs of liquid may then be accelerated towards each other or against the piping structure. The possibility of this phenomenon is of concern to us because it may become a dominant phenomenon and change the character of the transient. In performing the code scaling, applicability, and uncertainty study (CSAU) on a SBLOCA scenario, we had to examine the possibility that the transient being analyzed could experience water-hammer and thus depart from the scope of the study. Two criteria for water-hammer initiation were investigated and tested using a RELAP5/MOD3 simulation of the transient. Our results indicated a very low likelihood of occurrence of the phenomenon. 8 refs., 6 figs

  1. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Park, Kyung Soo; Chu, Won Ho [Korea Maritime Univ., Jinhae (Korea, Republic of)

    2003-03-15

    The objectives of the present work are to carry out technical evaluation and review of CANDU safety analysis methods in order to assist development of performance evaluation methods and review guides for CANDU ECCS. The applicability of PWR ECCS analysis models are examined and it suggests that unique data or models for CANDU are required for the following phenomena: break characteristics and flow, frictional pressure drop, post-CHF heat transfer correlations, core flow distribution during blowdown, containment pressure, and reflux rate. For safety analysis of CANDU, conservative analysis or best estimate analysis can be used. The main advantage of BE analysis is a more realistic prediction of margins to acceptance criteria. The expectation is that margins demonstrated with BE methods would be larger that when a conservative approach is applied. Some outstanding safety analysis issues can be resolved by demonstration that accident consequences are more benign than previously predicted. Success criteria for analysis and review of Large LOCA can be developed by top-down approach. The highest-level success criteria can be extracted from C-6 and from them, the lower level criteria can be developed step-by-step, in a logical fashion. The overall objectives for analysis and review are to verify radiological consequences and frequency are met.

  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. 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

  4. 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)

  5. 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)

  6. 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

  7. 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.

  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 (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

  9. 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.

  10. 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)

  11. 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.

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. 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.

  17. 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)

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. 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

  9. 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

  10. Two-phase flow dynamics in ECC

    International Nuclear Information System (INIS)

    Albraaten, P.J.

    1981-07-01

    The present report summarizes the achievements within the project ''Two-phase Systems and ECC''. The results during 1978 - 1980 are accounted for in brief as they have been documented in earlier reports. The results during the first half of 1981 are accounted for in greater detail. They contain a new model for the Basset force and test runs with this model using the test code RISQUE. Furthermore, test runs have been performed with TRAC-PD2 MOD 1. This code was implemented on Edwards Pipe Blowdown experiment (a standard test case) and UC-Berkeley Reflooding experiment (a non-standard test case.) (Auth.)

  11. Noise-based Stego-ECC

    Directory of Open Access Journals (Sweden)

    Rahardjo Budi

    2014-03-01

    Full Text Available A novel method of inserting noise into stream of ciphered text is proposed. The goal of inserting noise is to increase the level of uncertainty, thus making it harder for an attacker to detect data and noise. This form of steganography is implemented using Elliptic Curve Cryptography (ECC. The process of embedding the noise to the message in the encryption process and removing the noise from the message in the decryption process is proposed in this work by modifying ElGamal to allow auto detection of data and noise.

  12. Coolability of severely degraded CANDU cores

    International Nuclear Information System (INIS)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Mijhawan, S.

    1995-07-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually re solidify. Thus, the calandria vessel would act inherently as a core-catcher as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author). 48 refs., 3 tabs., 18 figs

  13. Coolability of severely degraded CANDU cores. Revised

    International Nuclear Information System (INIS)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S.

    1996-01-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)

  14. 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

  15. Shear behavior of reinforced Engineered Cementitious Composites (ECC) beams

    DEFF Research Database (Denmark)

    Paegle, Ieva; Fischer, Gregor

    2010-01-01

    This paper describes an experimental investigation of the shear behavior of beams consisting of steel reinforced Engineered Cementitious Composites (ECC). Based on the strain hardening and multiple cracking behavior of ECC, this study investigates the extent to which ECC can improve the shear...... capacity of beams loaded primarily in shear and if ECC can partially or fully replace the conventional transverse steel reinforcement in beams. However, there is a lack of understanding of how the fibers affect the shear carrying capacity and deformation behavior of structural members if used either...

  16. Final environmental statement concerning proposed rule making action: acceptance criteria for emergency core cooling systems for light-water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    1973-01-01

    Information on the ECCS requirements for PWR and BWR reactors are presented concerning environmental impact, alternatives, adverse environmental effects, relationship between short-term uses of the environment and long-term productivity, irreversible and irretrievable commitment of resources, cost-benefit analysis, and discussion of significant comments received on the draft environmental statement. (U.S.)

  17. 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).

  18. 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

  19. 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

  20. 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

  1. A LOCA analysis for AHWR caused by ECCS header rupture

    International Nuclear Information System (INIS)

    Chatterjee, B.; Gawai, Amol; Gupta, S.K.; Kushwaha, H.S.

    2000-01-01

    Loss of coolant accident (LOCA) analyses for the proposed 750 MWth Advanced Heavy Water Reactor (AHWR), initiated by the rupture of 8 inch NB ECCS header has been carried out. This paper narrates the description of AHWR and associated ECCS, postulated scenario with which the analyses is carried out, results, discussion and conclusion

  2. 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.

  3. 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.

  4. ECCE Toolkit: Prototyping Sensor-Based Interaction

    Directory of Open Access Journals (Sweden)

    Andrea Bellucci

    2017-02-01

    Full Text Available Building and exploring physical user interfaces requires high technical skills and hours of specialized work. The behavior of multiple devices with heterogeneous input/output channels and connectivity has to be programmed in a context where not only the software interface matters, but also the hardware components are critical (e.g., sensors and actuators. Prototyping physical interaction is hindered by the challenges of: (1 programming interactions among physical sensors/actuators and digital interfaces; (2 implementing functionality for different platforms in different programming languages; and (3 building custom electronic-incorporated objects. We present ECCE (Entities, Components, Couplings and Ecosystems, a toolkit for non-programmers that copes with these issues by abstracting from low-level implementations, thus lowering the complexity of prototyping small-scale, sensor-based physical interfaces to support the design process. A user evaluation provides insights and use cases of the kind of applications that can be developed with the toolkit.

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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

  11. EFFECT OF ACTIVE COOLING AND α-2 ADRENOCEPTOR ANTAGONISM ON CORE TEMPERATURE IN ANESTHETIZED BROWN BEARS (URSUS ARCTOS).

    Science.gov (United States)

    Ozeki, Larissa Mourad; Caulkett, Nigel; Stenhouse, Gordon; Arnemo, Jon M; Fahlman, Åsa

    2015-06-01

    Hyperthermia is a common complication during anesthesia of bears, and it can be life threatening. The objective of this study was to evaluate the effectiveness of active cooling on core body temperature for treatment of hyperthermia in anesthetized brown bears (Ursus arctos). In addition, body temperature after reversal with atipamezole was also evaluated. Twenty-five adult and subadult brown bears were captured with a combination of zolazepam-tiletamine and xylazine or medetomidine. A core temperature capsule was inserted into the bears' stomach or 15 cm into their rectum or a combination of both. In six bears with gastric temperatures≥40.0°C, an active cooling protocol was performed, and the temperature change over 30 min was analyzed. The cooling protocol consisted of enemas with 2 L of water at approximately 5°C/100 kg of body weight every 10 min, 1 L of intravenous fluids at ambient temperature, water or snow on the paws or the inguinal area, intranasal oxygen supplementation, and removing the bear from direct sunlight or providing shade. Nine bears with body temperature>39.0°C that were not cooled served as control for the treated animals. Their body temperatures were recorded for 30 min, prior to administration of reversal. At the end of the anesthetic procedure, all bears received an intramuscular dose of atipamezole. In 10 bears, deep rectal temperature change over 30 min after administration of atipamezole was evaluated. The active cooling protocol used in hyperthermic bears significantly decreased their body temperatures within 10 min, and it produced a significantly greater decrease in their temperature than that recorded in the control group.

  12. Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Feldman, E.M.; Collins, B.L.; Sackett, K.E.

    1977-02-01

    Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

  13. Limits on turbulent propagation of energy in cool-core clusters of galaxies

    Science.gov (United States)

    Bambic, C. J.; Pinto, C.; Fabian, A. C.; Sanders, J.; Reynolds, C. S.

    2018-07-01

    We place constraints on the propagation velocity of bulk turbulence within the intracluster medium of three clusters and an elliptical galaxy. Using Reflection Grating Spectrometer measurements of turbulent line broadening, we show that for these clusters, the 90 per cent upper limit on turbulent velocities when accounting for instrumental broadening is too low to propagate energy radially to the cooling radius of the clusters within the required cooling time. In this way, we extend previous Hitomi-based analysis on the Perseus cluster to more clusters, with the intention of applying these results to a future, more extensive catalogue. These results constrain models of turbulent heating in active galactic nucleus feedback by requiring a mechanism which can not only provide sufficient energy to offset radiative cooling but also resupply that energy rapidly enough to balance cooling at each cluster radius.

  14. 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

  15. 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

  16. 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.)

  17. Superficial cooling does not decrease core body temperature before, during, or after exercise in an American football uniform.

    Science.gov (United States)

    Lopez, Rebecca M; Eberman, Lindsey E; Cleary, Michelle A

    2012-12-01

    The purpose of this study was to identify the effects of superficial cooling on thermoregulatory responses while exercising in a hot humid environment while wearing an American football uniform. Nine male and female subjects wore a superficial cooling garment while in a cooling (CS) experimental condition or a no cooling (NCS) control condition during an exercise task consisting of warm-up (WU), exercise (EX), and recovery (R). The exercise task simulated an American football conditioning session with subjects wearing a full American football uniform and performing anaerobic and aerobic exercises in a hot humid environment. Subjects were allowed to drink water ad libitum during rest breaks. During the WU, EX, and R periods, core body temperature (T(c)) was measured to assess the effect of the cooling garment. Neither baseline resting before warm-up T(c) nor after warm-up T(c) was significantly different between trials. No significant differences in exercise T(c) between conditions were found. Time to return to baseline T(c) revealed no significant differences between the experimental and control conditions. The authors found that the volume of fluid consumed was 34% less in the experimental condition (711.1 ± 188.0 ml) compared with the control condition (1,077.8 ± 204.8 ml). The findings indicate that the cooling garment was not effective in blunting the rise in T(c) during warm-up, attenuating a rise in T(c) during intermittent exercise, or in increasing a return to baseline T(c) during a resting recovery period in a hot humid environment while wearing an American football uniform.

  18. Preliminary test conditions for KNGR SBLOCA DVI ECCS performance test

    International Nuclear Information System (INIS)

    Bae, Kyoo Whan; Song, Jin Ho; Chung, Young Jong; Sim, Suk Ku; Park, Jong Kyun

    1999-03-01

    The Korean Next Generation Reactor (KNGR) adopts 4-train Direct Vessel Injection (DVI) configuration and injects the safety injection water directly into the downcomer through the 8.5'' DVI nozzle. Thus, the thermal hydraulic phenomena such as ECCS mixing and bypass are expected to be different from those observed in the cold leg injection. In order to investigate the realistic injection phenomena and modify the analysis code developed in the basis of cold leg injection, thermal hydraulic test with the performance evaluation is required. Preliminarily, the sequence of events and major thermal hydraulic phenomena during the small break LOCA for KNGR are identified from the analysis results calculated by the CEFLASH-4AS/REM. It is shown from the analysis results that the major transient behaviors including the core mixture level are largely affected by the downcomer modeling. Therefore, to investigate the proper thermal hydraulic phenomena occurring in the downcomer with limited budget and time, the separate effects test focusing on this region is considered to be effective and the conceptual test facility based on this recommended. For this test facility the test initial and boundary conditions are developed using the CEFLASH-4AS/REM analysis results that will be used as input for the preliminary test requirements. The final test requirements will be developed through the further discussions with the test performance group. (Author). 10 refs., 18 tabs., 4 figs

  19. Prediction, analysis and solution of flow inversion phenomenon in a typical MTR reactor with upward core cooling

    International Nuclear Information System (INIS)

    El-Morshedy, Salah El-Din

    2010-01-01

    Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling under a hypothetical case of loss of off-site power. The flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. The reactor simulation under loss of off-site power is performed for two cases namely; two-flap valves open and one flap-valve fails to open. The model results for the flow inversion phenomenon prediction is analyzed and a solution of the problem is suggested. (orig.)

  20. Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

    Energy Technology Data Exchange (ETDEWEB)

    Fabbris, Olivier [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Dardour, Saied, E-mail: saied.dardour@cea.fr [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France); Blaise, Patrick [CEA DEN/DER/SPEX, 13108 Saint-Paul-Lez-Durance (France); Ferrasse, Jean-Henry [Aix-Marseille Université, CNRS, ECM, M2P2 UMR 7340, 13451 Marseille (France); Saez, Manuel [CEA DEN/DER/SESI, 13108 Saint-Paul-Lez-Durance (France)

    2016-08-15

    Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

  1. Surrogates based multi-criteria predesign methodology of Sodium-cooled Fast Reactor cores – Application to CFV-like cores

    International Nuclear Information System (INIS)

    Fabbris, Olivier; Dardour, Saied; Blaise, Patrick; Ferrasse, Jean-Henry; Saez, Manuel

    2016-01-01

    Highlights: • We developed an ERANOS calculation scheme to evaluate the neutronics of CFV cores. • We used this scheme to simulate a number if cores within a predefined study space. • Simulation results were used to build surrogate models describing CFV neutronics. • These models were used to carry on global sensitivity analyses. • The methodology helped identify the most important core design parameters. - Abstract: The Sodium-cooled Fast Reactor (SFR) core predesign process is commonly realized on the basis of expert advices and local parametric studies. As such, in-deep knowledge of physical phenomena avoids an important number of expensive simulations. However, the study space is explored only partially. To ease the computational burden metamodels, or surrogate models, can be used, to quickly evaluate the performances of a wide set of different cores, individually defined by a set of parameters (pellet diameter, fissile height…), in the study space. This paper presents the development of a simplified neutronics ERANOS reference core calculation scheme that is then implemented in the construction of the Design of Experiment (DOE) database. The surrogate models for SFR CFV-like cores performances are developed, biases and uncertainties are quantified against the CFV-v1 version. Global Sensitivity Analysis also allowed highlighting antagonist performances for the design and to propose two alternative core configurations. A broadened application of the method with an optimization of a CFV-like core is also detailed. The Pareto front of the seven selected performance parameters has been studied using eleven surrogate models, based on Artificial Neural Network (ANN). The optimization demonstrates that the CFV-v1, designed using Best Estimate codes, under given performance constraints, is Pareto optimal: no other configuration is highlighted from the Multi-Objective Optimization (MOO) study. Further MOO analysis, including a specific study on impact of new

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

  5. Experimental and numerical simulation of passive decay heat removal by sump cooling after core melt down

    International Nuclear Information System (INIS)

    Knebel, J.U.; Mueller, U.

    1997-01-01

    This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The results are applied to the SUCO program that experimentally and numerically investigates 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 results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. (author)

  6. Experimental and numerical simulation of passive decay heat removal by sump cooling after core melt down

    Energy Technology Data Exchange (ETDEWEB)

    Knebel, J.U.; Mueller, U. [Forschungszentrum Karlsruhe - Technik und Umwelt Inst. fuer Angewandte Thermo- und Fluiddynamik (IATF), Karlsruhe (Germany)

    1997-12-31

    This article presents the basic physical phenomena and scaling criteria of passive decay heat removal from a large coolant pool by single-phase natural circulation. The physical significance of the dimensionless similarity groups derived is evaluated. The results are applied to the SUCO program that experimentally and numerically investigates 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 results of temperature and velocity measurements in the 1:20 linearly scaled SUCOS-2D test facility. The experiments are backed up by numerical calculations using the commercial software Fluent. Finally, using the similarity analysis from above, the experimental results of the model geometry are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. (author)

  7. Study of the core compaction effects and its monitoring in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zylbersztejn, F.

    2012-01-01

    Conclusions: • On calculation of reactivity impacts of core compaction/flowering: → Upper bound of the reactivity coefficients for each type of deformation; → Uniform compaction model: significant reactivity impact; Circular symmetric model: small reactivity impact. • On the visibility of these phenomena by the neutron detectors: → The direct monitoring of the core compaction by neutron detector in the BCC is not possible. (the identification that the reactivity perturbations observed are due to variation of the core geometry). Perspectives of solutions: → Improved core design: reducing the effects. → Physical improvements: Steel resistance to deformations (irradiation, flexion); Direct devices: core constraint (prevents deformations). → Additional calculations: Considering more localized deformations; Advanced monitoring with neutron noise (in progress)

  8. A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

    International Nuclear Information System (INIS)

    Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

    2009-01-01

    The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

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

    International Nuclear Information System (INIS)

    1991-03-01

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

  10. 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

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

    International Nuclear Information System (INIS)

    Kaminaga, Masanori; Watanabe, Shukichi; Ando, Hiroei; Sudo, Yukio; Ikawa, Hiromasa.

    1987-03-01

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

  12. 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)

  13. 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)

  14. Materials problems related to the core catcher of sodium cooled reactors

    International Nuclear Information System (INIS)

    Goetzmann, O.

    1975-05-01

    There are in principal two possible solutions for the external core catcher as far as materials are concerned. 1) A barrier consisting of a material with a high melting point, 2) a tray of comparatively low melting material with a high solubility for the fuel. In case of the first concept one has to look for materials whose melting temperatures are above the temperature of the molten core. Based on metallurgical reasons it seems very likely that the molten core does not exceed a temperature in the range between 2,500 and 2,800 0 C. Due to the compatibility situation with the molten core only a few high melting oxides will be suitable as liner materials for a core catcher. In the second case basalt or concrete, if free of water and lime, are suitable materials. Graphite is a high melting material, however, due to its behaviour with the molten core it should be listed under the second group. By the reaction of graphite with the core materials the melt can be kept liquid down to temperatures of around 1,100 0 C. The evolution of CO by this reaction should be supportable. It is an endothermal reaction. Experiments on the behaviour of core catcher materials have shown that sodium is capable of penetrating into sintered bodies of UO 2 with densities of 90% TD at temperatures higher than 200 0 C. This may lead to the desintegration of these bodies. The exposure to moist air has not done much harm to UO 2 pellets of densities from 80 to 90% TD. Even after one year of exposure, swelling or desintegration could not be observed. Sodium is also capable of penetrating into bodies of synthetic carbon and graphite. Only well graphitized material will not be destroyed. (orig.) [de

  15. 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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  17. A method and programme (BREACH) for predicting the flow distribution in water cooled reactor cores

    International Nuclear Information System (INIS)

    Randles, J.; Roberts, H.A.

    1961-03-01

    The method presented here of evaluating the flow rate in individual reactor channels may be applied to any type of water cooled reactor in which boiling occurs The flow distribution is calculated with the aid of a MERCURY autocode programme, BREACH, which is described in detail. This programme computes the steady state longitudinal void distribution and pressure drop in a single channel on the basis of the homogeneous model of two phase flow. (author)

  18. 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

  19. A method and programme (BREACH) for predicting the flow distribution in water cooled reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J; Roberts, H A [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1961-03-15

    The method presented here of evaluating the flow rate in individual reactor channels may be applied to any type of water cooled reactor in which boiling occurs The flow distribution is calculated with the aid of a MERCURY autocode programme, BREACH, which is described in detail. This programme computes the steady state longitudinal void distribution and pressure drop in a single channel on the basis of the homogeneous model of two phase flow. (author)

  20. 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)

  1. 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

  2. Melt cooling by bottom flooding. The COMET core-catcher concept

    International Nuclear Information System (INIS)

    Foit, Jerzy Jan; Alsmeyer, Hans; Tromm, Walter; Buerger, Manfred; Journeau, Christophe

    2009-01-01

    The COMET concept has been developed to cool an ex-vessel corium melt in case of a hypothetical severe accident leading to vessel melt-through. After erosion of a sacrificial concrete layer the melt is passively flooded by bottom injection of coolant water. The open porosities and large surface that are generated during melt solidification form a porous permeable structure that is permanently filled with the evaporating water and thus allows an efficient short-term as well as long-term removal of the decay heat. The advantages of this concept are the fast cool-down and complete solidification of the melt within less than one hour typically. This stops further release of fission products from the corium. A drawback may be the fast release of steam during the quenching process. Several experimental series have been performed by FZK (Germany) to test and optimise the functionality of the different variants of the COMET concept. Thermite generated melts of iron and aluminium oxide were used. The large scale COMET-H test series with sustained inductive heating includes nine experiments performed with an array of water injection channels embedded in a sacrificial concrete layer. Variation of the water inlet pressure and melt height showed that melts up to 50 cm height can be safely cooled with an overpressure of the coolant water of 0.2 bar. The CometPC concept is based on cooling by flooding the melt from the bottom through layers of porous, water filled concrete. The third variant of the COMET design, CometPCA, uses a layer of porous, water filled concrete CometPCA from which flow channels protrude into the layer of sacrificial concrete. This modified concept combines the advantages of the original COMET concept with flow channels and the high resistance of a water-filled porous concrete layer against downward melt attack. Four large scale CometPCA experiments (FZK, Germany) have demonstrated an efficient cooling of melts up to 50 cm height using the recommended water

  3. Status of experimental verification of ECCS efficiency

    International Nuclear Information System (INIS)

    Hein, D.; Watzinger, H.

    1978-01-01

    For the emergency cooling system of KWU pressurized water reactors with combined hot and cold leg injection an outline is given of the status of experiments designed to prove the efficiency of the emergency cooling system. This proof has been established by basic investigations which clarify the physical processes, by ''separate effects tests'' to derive and check correlations, and finally by investigations on the PKL test facility, in which a 1300 MWe pressurized water reactor including the primary circuts is simulated. These ''system effects tests'' are used to verify computer codes which are ultimately used to make predictions for the reactor. (author)

  4. IEA-R1 reactor core simulation with RELAP5 code

    International Nuclear Information System (INIS)

    Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

    2005-01-01

    This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

  5. Parametric study of the potential for BWR ECCS strainer blockage due to LOCA generated debris. Final report

    International Nuclear Information System (INIS)

    Zigler, G.; Brideau, J.; Rao, D.V.; Shaffer, C.; Souto, F.; Thomas, W.

    1995-10-01

    This report documents a plant-specific study for a BWR/4 with a Mark I containment that evaluated the potential for LOCA generated debris and the probability of losing long term recirculation capability due ECCS pump suction strainer blockage. The major elements of this study were: (1) acquisition of detailed piping layouts and installed insulation details for a reference BWR; (2) analysis of plant specific piping weld failure probabilities to estimate the LOCA frequency; (3) development of an insulation and other debris generation and drywell transport models for the reference BWR; (4) modeling of debris transport in the suppression pool; (5) development of strainer blockage head loss models for estimating loss of NPSH margin; (6) estimation of core damage frequency attributable to loss of ECCS recirculation capability following a LOCA. Elements 2 through 5 were combined into a computer code, BLOCKAGE 2.3. A point estimate of overall DEGB pipe break frequency (per Rx-year) of 1.59E-04 was calculated for the reference plant, with a corresponding overall ECCS loss of NPSH frequency (per Rx-year) of 1.58E-04. The calculated point estimate of core damage frequency (per Rx-year) due to blockage related accident sequences for the reference BWR ranged from 4.2E-06 to 2.5E-05. The results of this study show that unacceptable strainer blockage and loss of NPSH margin can occur within the first few minutes after ECCS pumps achieve maximum flows when the ECCS strainers are exposed to LOCA generated fibrous debris in the presence of particulates (sludge, paint chips, concrete dust). Generic or unconditional extrapolation of these reference plant calculated results should not be undertaken

  6. 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.

  7. Whole Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor Considering the Gamma Energy Transport

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Back, Min Ho; Park, Won Seok; Kim, Sang Ji

    2012-01-01

    Since a fuel cladding failure is the most important parameter in a core thermal-hydraulic design, the conceptual design stage only involves fuel assemblies. However, although non-fuel assemblies such as control rod, reflector, and B4C generate a relatively smaller thermal power compared to fuel assemblies, they also require independent flow allocation to properly cool down each assembly. The thermal power in non-fuel assemblies is produced from both neutron and gamma energy, and thus the core thermal-hydraulic design including non-fuel assemblies should consider an energy redistribution by the gamma energy transport. To design non-fuel assemblies, the design-limiting parameters should be determined considering the thermal failure modes. While fuel assemblies set a limiting factor with cladding creep temperature to prevent a fission product ejection from the fuel rods, non-fuel assemblies restrict their outlet temperature to minimize thermally induced stress on the upper internal structure (UIS). This work employs a heat generation distribution reflecting both neutron and gamma transport. The whole core thermal-hydraulic design including fuel and non-fuel assemblies is then conducted using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. The other procedures follow from the previous conceptual design

  8. 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

  9. 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.

  10. LWR fuel cladding deformation in a LOCA and its interaction with the emergency core cooling

    International Nuclear Information System (INIS)

    Erbacher, F.J.

    1982-01-01

    The paper summarizes research results of out-of-pile burst tests, in-pile bursts tests, out-of-pile flooding tests and modeling work on fuel behavior in a LOCA performed at KfK: The dominant phenomena of the cladding deformation and failure have been clarified by experiments and can be modeled by computer codes. The burst and flooding tests performed up to now suggest that the coolability of the core under LOCA conditions can be maintained. (orig.) [de

  11. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Min; Sabundjian, Gaianê, E-mail: smlee@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  12. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    International Nuclear Information System (INIS)

    Lee, Seung Min; Sabundjian, Gaianê

    2017-01-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  13. 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

  14. Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Raza, S.H.

    1993-12-01

    Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs

  15. MIF-SCD computer code for thermal hydraulic calculation of supercritical water cooled reactor core

    International Nuclear Information System (INIS)

    Galina P Bogoslovskaia; Alexander A Karpenko; Pavel L Kirillov; Alexander P Sorokin

    2005-01-01

    Full text of publication follows: Supercritical pressure power plants constitute the basis of heat power engineering in many countries to day. Starting from a long-standing experience of their operation, it is proposed to develop a new type of fast breeder reactor cooled by supercritical water, which enables the economical indices of NPP to be substantially improved. In the Thermophysical Department of SSC RF-IPPE, an attempt is made to provide thermal-hydraulic validation of the reactor under discussion. The paper presents the results of analysis of the thermal-hydraulic characteristics of fuel subassemblies cooled by supercritical water based on subchannel analysis. Modification of subchannel code MIF - MIF-SCD Code - developed in the SSC RF IPPE is designed as block code and permits one to calculate the coolant temperature and velocity distributions in fuel subassembly channels, the temperature of fuel pin claddings and fuel subassembly wrapper under conditions of irregular geometry and non-uniform axial and radial power generation. The thermal hydraulics under supercritical pressure of water exhibits such peculiarities as abrupt variation of the thermal physical properties in the range of pseudo-critical temperature, the absence of such phenomenon as the critical heat flux which can lead to fuel element burnout in WWERs. As compared with subchannel code for light water, in order to take account of the variation of the coolant properties versus temperature in more detail, a block for evaluating the thermal physical properties of supercritical water versus the local coolant temperature in the fuel subassembly channels was added. The peculiarities of the geometry and power generation in the fuel subassembly of the supercritical reactor are considered as well in special blocks. The results of calculations have shown that considerable preheating of supercritical coolant (several hundreds degrees) can occur in the fuel subassembly. The test calculations according to

  16. 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

  17. 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)

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

    International Nuclear Information System (INIS)

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

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

  20. 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)

  1. Application of Engineered Cementitious Composites (ECC) in modular floor panels

    DEFF Research Database (Denmark)

    Lárusson, Lárus Helgi; Fischer, Gregor; Jönsson, Jeppe

    2008-01-01

    This paper describes the design, manufacturing, and structural behavior of a prefabricated floor panel consisting of a modular assembly of a thin-walled ECC slab and steel truss girders. The features of this composite structure include light weight, the modular manufacturing process...... concept introduced in this paper aims at improvements in the manufacturing process of the panels by casting the ECC slab separately and subsequently joining it with the steel trusses. The focus of this paper is on design and manufacturing of a prototype modular panel and on its structural behavior under...

  2. 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)

  3. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

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

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

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

  5. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  6. 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.

  7. Pre-conceptual core design of a small modular fast reactor cooled by supercritical CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Baolin; Cao, Liangzhi; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); Yuan, Xianbao, E-mail: ztsbaby@163.com [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); College of Mechanical & Power Engineering, China Three Gorges University, No 8, Daxue Road, Yichang 443002, Hubei (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China)

    2016-04-15

    Abstracts: A Small Modular fast reactor cooled by Supercritical CO{sub 2} (SMoSC) is pre-conceptually designed through three-dimensional coupled neutronics/thermal-hydraulics analysis. The power rating of the SMoSC is designed to be 300 MW{sub th} to meet the energy demand of small electrical grids. The excellent thermal properties of supercritical CO{sub 2} (S-CO{sub 2}) are employed to obtain a high thermal efficiency of about 40% with an electric output of 120 MWe. MOX fuel is utilized in the core design to improve fuel efficiency. The tube-in-duct (TID) assembly is applied to get lower coolant volume fraction and reduce the positive coolant void reactivity. According to the coupled neutronics/thermal-hydraulics calculations, the coolant void reactivity is kept negative throughout the whole core life. With a specific power density of 9.6 kW/kg and an average discharge burnup of 70.1 GWd/tHM, the SmoSC can be operated for 20 Effective Full Power Years (EFPYs) without refueling.

  8. 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

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

    International Nuclear Information System (INIS)

    DeVault, G.P.; Bell, C.R.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, Liancheng; Zhang, Bin

    2014-01-01

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

  11. Exxon Nuclear Company ECCS evaluation of a 2-loop Westinghouse PWR with dry containment using the ENC WREM-II ECCS model. Large break example problem

    International Nuclear Information System (INIS)

    Krajicek, J.E.

    1977-01-01

    This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77 for heatup analysis

  12. Calculations of combined radiation and convection heat transfer in rod bundles under emergency cooling conditions

    International Nuclear Information System (INIS)

    Sun, K.H.; Gonzalez-Santalo, J.M.; Tien, C.L.

    1976-01-01

    A model has been developed to calculate the heat transfer coefficients from the fuel rods to the steam-droplet mixture typical of Boiling Water Reactors under Emergency Core Cooling System (ECCS) operation conditions during a postulated loss-of-coolant accident. The model includes the heat transfer by convection to the vapor, the radiation from the surfaces to both the water droplets and the vapor, and the effects of droplet evaporation. The combined convection and radiation heat transfer coefficient can be evaluated with respect to the characteristic droplet size. Calculations of the heat transfer coefficient based on the droplet sizes obtained from the existing literature are consistent with those determined empirically from the Full-Length-Emergency-Cooling-Heat-Transfer (FLECHT) program. The present model can also be used to assess the effects of geometrical distortions (or deviations from nominal dimensions) on the heat transfer to the cooling medium in a rod bundle

  13. Operating experience of natural circulation core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    Kullberg, C.; Jones, K.; Heath, C.

    1993-01-01

    General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

  14. Modeling of ECC materials using numerical formulations based on plasticity

    DEFF Research Database (Denmark)

    Dick-Nielsen, Lars; Stang, Henrik; Poulsen, Peter Noe

    2006-01-01

    scale it is shown that the cohesive law for a unidirectional fiber reinforced cementitious composite can be found through superposition of the cohesive law for mortar and the fiber bridging curve. On the meso scale I it is shown that the maximum crack opening observed during crack propagation in ECC...

  15. ECCS evaluation of B and W's 205-FA NSS

    International Nuclear Information System (INIS)

    Lowe, R.J.; Anderson, G.E. Jr.; Dunn, B.M.

    1975-06-01

    The effectiveness of the ECCS for B and W's 205-fuel assembly plants is evaluated and shown to meet all the requirements of 10 CFR 50.46. The results of various sensitivity studies, a spectrum of breaks, and an analysis to determine allowable linear heat rates under 10 CFR 50.46 are presented. (14 references) (U.S.)

  16. 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

  17. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R.

    2015-09-01

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm 2 and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  18. Non-equilibrium effects of core-cooling and time-dependent internal heating on mantle flush events

    Directory of Open Access Journals (Sweden)

    D. A. Yuen

    1995-01-01

    Full Text Available We have examined the non-equilibrium effects of core-cooling and time-dependent internal-heating on the thermal evolution of the Earth's mantle and on mantle flush events caused by the two major phase transitions. Both two- and three-dimensional models have been employed. The mantle viscosity responds to the secular cooling through changes in the averaged temperature field. A viscosity which decreases algebraically with the average temperature has been considered. The time-dependent internal-heating is prescribed to decrease exponentially with a single decay time. We have studied the thermal histories with initial Rayleigh numbers between 2 x 107 and 108 . Flush events, driven by the non-equilibrium forcings, are much more dramatic than those produced by the equilibrium boundary conditions and constant internal heating. Multiple flush events are found under non-equilibrium conditions in which there is very little internal heating or very fast decay rates of internal-heating. Otherwise, the flush events take place in a relatively continuous fashion. Prior to massive flush events small-scale percolative structures appear in the 3D temperature fields. Time-dependent signatures, such as the surface heat flux, also exhibits high frequency oscillatory patterns prior to massive flush events. These two observations suggest that the flush event may be a self-organized critical phenomenon. The Nusselt number as a function of the time-varying Ra does not follow the Nusselt vs. Rayleigh number power-law relationship based on equilibrium (constant temperature boundary conditions. Instead Nu(t may vary non-monotonically with time because of the mantle flush events. Convective processes in the mantle operate quite differently under non-equilibrium conditions from its behaviour under the usual equilibrium situations.

  19. Regulatory guide in support of ECCS rule revision

    International Nuclear Information System (INIS)

    Tovmassian, H.S.

    1987-01-01

    The US Nuclear Regulatory Commission staff is proposing to amend 10 CFR 50.46 and Appendix K to allow licensees to use best estimate calculations to estimate emergency core cooling system performance. This estimate in conjunction with an estimate of the uncertainty in the calculation would then be used to assure that the licensing limits set forth in 10 CFR 50.46(b) are not exceeded. The NRC staff has prepared a draft regulatory guide to assist licensees and applicants in complying with these proposed amendments. This paper sets forth the objectives of this regulatory guide, the approach taken, the difficulties encountered, and the current status of this effort

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

    International Nuclear Information System (INIS)

    Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn; Morrow, Charles; Osborn, Douglas; Gauntt, Randall O.

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Aghoyeh, Reza Gholizadeh [School of Research and Development of Nuclear Reactors and Accelerators, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Khalafi, Hossein, E-mail: hkhalafi@aeoi.org.i [School of Research and Development of Nuclear Reactors and Accelerators, Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran (AEOI), P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of)

    2011-03-15

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

  3. Risk-based assessment of the allowable outage times for the unit 1 leningrad nuclear power plant ECCS components

    International Nuclear Information System (INIS)

    Koukhar, Sergey; Vinnikov, Bronislav

    2009-01-01

    Present paper describes a method for risk - informed assessment of the Allowable Outage Times (AOTs). The AOT is the time, when components of a safety system allowed to be out of service during power operation or during shutdown operation off a plant. If the components are not restored during the time, the plant in operation must be shut down or the plant in a given shutdown mode has to go to safer shutdown mode. Application of the method is also provided for the equipment of the Unit 1 Leningrad NPP ECCS components. For solution of the problem it is necessary to carry out two series of computations using a Living PSA model, level 1. In the first series of the computations the core damage frequency (CDFb) for the base configuration of the plant is determined (there is no equipment out of service). Here the symbol 'b' means the base configuration of a plant. In the second series of the computations the core damage frequency (CDFi) for the configuration of the plant with the component (which is out of service) is calculated. That is here CDFi is determined for the failure probability of the component equal to 1.0 (component 'i' is unavailable). Then it is necessary to determine so called Risk Increase Factor (RIF) using the following ratio: RIFi = CDFi / CDFb. At last the AOT is calculated with the help of the ratio: AOTi = Tppr / RIFi, where Tppr is a period of time between two Planned Preventive Repairs (PPRs). 1. Using the risk based approach the AOTs were calculated for a set of the components of the Unit 1 Leningrad NPP ECCS components. 2. The main conclusion from the analysis is that the current deterministic AOTs for the ECCS components are conservative and should be extended. 3. The risk based extension of the AOTs for the ECCS components can prevent the Unit 1 Leningrad NPP to enter into the operating modes with increased risk. (author)

  4. Experimental investigation of material chemical effects on emergency core cooling pump suction filter performance after loss of coolant accident

    International Nuclear Information System (INIS)

    Park, Jong Woon; Park, Byung Gi; Kim, Chang Hyun

    2009-01-01

    Integral tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for 30 days after a loss of coolant accident. A test rig with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber and the amounts of reactor building materials are scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of calcium-silicate, reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON TM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

  5. VizieR Online Data Catalog: Cool-core clusters with Chandra obs. (Andrade-Santos+, 2017)

    Science.gov (United States)

    Andrade-Santos, F.; Jones, C.; Forman, W. R.; Lovisari, L.; Vikhlinin, A.; van Weeren, R. J.; Murray, S. S.; Arnaud, M.; Pratt, G. W.; Democles, J.; Kraft, R.; Mazzotta, P.; Bohringer, H.; Chon, G.; Giacintucci, S.; Clarke, T. E.; Borgani, S.; David, L.; Douspis, M.; Pointecouteau, E.; Dahle, H.; Brown, S.; Aghanim, N.; Rasia, E.

    2018-02-01

    The main goal of this work is to compare the fraction of cool-core (CC) clusters in X-ray-selected and SZ-selected samples. The first catalog of 189 SZ clusters detected by the Planck mission was released in early 2011 (Planck Collaboration 2011, VIII/88/esz). A Chandra XVP (X-ray Visionary Program--PI: Jones) and HRC Guaranteed Time Observations (PI: Murray) combined to form the Chandra-Planck Legacy Program for Massive Clusters of Galaxies. For each of the 164 ESZ Planck clusters at z<=0.35, we obtained Chandra exposures sufficient to collect at least 10000 source counts. The X-ray sample used here is an extension of the Voevodkin & Vikhlinin (2004ApJ...601..610V) sample. This sample contains 100 clusters and has an effective redshift depth of z<0.3. All have Chandra observations. Of the 100 X-ray-selected clusters, 49 are also in the ESZ sample, and 47 are in the HIFLUGCS (Reiprich & Boehringer 2002ApJ...567..716R) catalog. (2 data files).

  6. 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

  7. Fuel and core design study of the sodium-cooled fast reactors. Studies on metallic fuel cores in the JFY2002

    International Nuclear Information System (INIS)

    Sugino, Kazuteru; Mizuno, Tomoyasu

    2003-06-01

    Based on the results obtained in the former feasibility study, the metallic fueled core of ordinary-type, that is, 2-region homogeneous core, has been established aiming at the improvement in the core performance, and subsequent comparison has been performed with the mixed oxide fueled core. Further, the attractive concept of the metallic fueled core of high outlet temperature has been constructed which has good nuclear features as a metallic fueled core and has identical outlet temperature to mixed oxide fuelled core. Following items have been found as a result of the investigation on the ordinary-type core. The metallic fueled core whose maximum fast neutron fluence (En>0.1MeV) is set identical (5x10 23 n/cm 2 ) to the mixed oxide fueled cores with core discharge burnup 150GWd/t has sufficient core performances as a metallic fueled core, e.g. higher breeding ratio and longer operation period compared with mixed oxide fueled cores, but the core discharge burnup is limited up to 100GWd/t. However effective discharge burnup including the contribution of the blanket region is comparative to mixed oxide cores under the same breeding ratio condition. In order to enlarge the core discharge burnup to 150GWd/t keeping the core performance identical to above mentioned core's, the irradiation deformation of structural material should be reduced to that of mixed oxide fueled cores. Further the maximum fast neutron fluence reaches to 7-8x10 23 n/cm 2 (En>0.1MeV). The investigations on the core of high outlet temperature have clarified following items. Even in the change of core regions by pin-diameter form 3-region to 2-region and in the limited maximum fuel pin diameter 8.5 mm, realization of the identical outlet/inlet temperatures to the mixed oxide cores (550/395degC) is feasible under the criteria of the maximum temperature 650degC at the inner surface of the cladding. The constructed core accommodates the targets of breeding ratio from about 1.0 to 1.2 only by adjusting

  8. LOFT ECC Pitot Tube and Thermocouple Rake Penetration thermal analysis

    International Nuclear Information System (INIS)

    Tolan, B.J.

    1977-01-01

    A thermal analysis of the LOFT ECC Pitot Tube and Thermocouple Rake Penetration was performed using COUPLE, a two-dimensional finite element computer code. Four transients which conservatively cover all transients the rake will be exposed to were included in this analysis in order to comply with the ASME Code Section III requirements. The transients conservatively cover hot and cold leg operation, and nuclear and nonnuclear operation. The four transients include the LOCE with ECC injection transient, the single control rod drop transient, the scram transient, and the heatup with 0 to 100% load change transient. Temperature distributions in the rake were obtained for each of the four transients and several plots of node temperatures vs. time are given

  9. An overview of the BWR ECCS strainer blockage issues

    International Nuclear Information System (INIS)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

    1996-01-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, open-quotes Containment Emergency Sump Performance,close quotes and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts

  10. Measurement and analysis of the ECC cusp structure

    Energy Technology Data Exchange (ETDEWEB)

    Pregliasco, R.G.; Garibotti, C.R.; Barrachina, R.O. (Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche)

    1994-03-28

    We measured the electron capture to the continuum (ECC) peak in the ionization of He atoms by the collision of 50, 100 and 200 keV H[sup +] and 100 keV amu[sup -1]He[sup 2+]. We exhaustively scanned the double differential cross section between the angles -1.5[sup o] to 10[sup o] and within the velocities (1 [+-] 0.15)[upsilon][sub i] where [upsilon][sub i] is the impact velocity. We propose a new method for analysing the ECC structure which is independent of the spectrometer transmission function modellization and enables a direct comparison with the available theories. Furthermore, we distinguish two regimes for the ionization process, depending on whether the electron velocity is larger or smaller than the projectile velocity. (Author).

  11. The microstructure of self-healed PVA ECC under wet and dry cycles

    Directory of Open Access Journals (Sweden)

    Jia Huan Yu

    2010-06-01

    Full Text Available Self-healing of ECC (Engineered Cementitious Composites subjected to cyclic wetting and drying regimes is investigated in this paper. ECC structures subjected to outdoor environmental conditions such as wind and rain runoff can be simulated by accelerated test method of wetting and drying cycles. Uniaxial tensile tests of ECC M45 and ECC 2.8FA specimen are conducted respectively. It is found that crack width of ECC 2.8FA is around 10 μm with increased amount of fly ash, while the crack width of ECC M45 is around 100 μm. New insights about the microstructure and chemical composition analysis of ECC specimens initially cracked to 2% strain and then self-healed under wet-dry cycles are presented.

  12. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    Höhne, Thomas; Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter

    2013-01-01

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  13. 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

  14. Simulation of a large break loss of coolant (LBLOCA), without actuation of the emergency injection systems (ECCS) for a BWR-5; Simulacion de un escenario de perdida de refrigerante grande (LBLOCA), sin actuacion de los sistemas de inyeccion de emergencia (ECCS) para un reactor BWR-5

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas V, J.; Mugica R, C. A.; Lopez M, R., E-mail: jaime.cardenas@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the analysis of scenario for the loss of coolant case was realized with break at the bottom of a recirculation loop of a BWR-5 with containment type Mark II and a thermal power of 2317 MWt considering that not have coolant injection. This in order to observe the speed of progression of the accident, the phenomenology of the scenario, the time to reach the limit pressure of containment venting and the amount of radionuclides released into the environment. This simulation was performed using the MELCOR code version 2.1. The scenario posits a break in one of the shear recirculation loops. The emergency core cooling system (ECCS) and the reactor core isolation cooling (Rcic) have not credit throughout the event, which allowed achieve greater severity on scenario. The venting of the primary containment was conducted via valve of 30 inches instead of the line of 24 inches of wet well, this in order to have a larger area of exhaust of fission products directly to the reactor building. The venting took place when the pressure in the primary containment reached the 4.5 kg/cm{sup 2} and remained open for the rest of the scenario to maximize the amount released of radionuclides to the atmosphere. The safety relief valves were considered functional they do not present mechanical failure or limit their ability to release pressure due to the large number of performances in safety mode. The results of the analysis covers about 48 hours, time at which the accident evolution was observed; behavior of level, pressure in the vessel and the fuel temperature profile was analyzed. For progression of the scenario outside the vessel, the pressure and temperature of the primary containment, level and temperature of the suppression pool, the hydrogen accumulation in the container and the radionuclides mass released into the atmosphere were analyzed. (Author)

  15. 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)

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

    International Nuclear Information System (INIS)

    McDonald, Michael

    2011-01-01

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

  17. Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

    1995-09-01

    Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

  18. Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow

    International Nuclear Information System (INIS)

    Whaley, R.L.; Sanders, J.P.

    1976-09-01

    A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection

  19. New ADAC centre cools and heats environmental-friendly. Concrete core tempering; Neue ADAC-Zentrale kuehlt und heizt umweltvertraeglich. Betonkerntemperierung

    Energy Technology Data Exchange (ETDEWEB)

    Anon,

    2010-12-15

    At present, the new centre of the General German Automobile Association (ADAC) arises in Munich's suburb Sendling/Westpark. In summer 2011, nearly 2,400 coworkers of ADAC so far distributed over six locations in Munich shall work together in the new office building. So that they feel well in this new building, Rehau AG (Rehau, Federal Republic of Germany) supplies an environmental-friendly heating and cooling of the building by means of concrete core tempering.

  20. 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

  1. Experiment data report for semiscale Mod-1 test S-04-2 (baseline ECC test)

    International Nuclear Information System (INIS)

    Crapo, H.S.; Collins, B.L.; Sackett, K.E.

    1976-09-01

    Recorded test data are presented for Test S-04-2 of the Semiscale Mod-1 Baseline ECC test series. This test is among Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor system. Test S-04-2 was conducted from an initial cold leg fluid temperature of 542 0 F and an initial pressure of 2,263 psia. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization and reflood transient using emergency core coolant injection parameters based on downcomer volume scaling. System flow was set to achieve a core fluid temperature differential of 66 0 F at a full core power of 1.6 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a flat radial power profile was used in the pressure vessel to simulate the effects of a nuclear core. During system depressurization, core power was reduced from the initial level of 1.6 MW to simulate the surface heat flux response of nuclear fuel rods until such sime that departure from nucleate boiling might occur. Blowdown to the pressure suppression system was accompanied by simulated emergency core coolant injection into both the intact and broken loops. Coolant injection was continued until test termination at 200 seconds after initiation of blowdown. The purpose of the report is to make available the uninterpreted data from Test S-04-2 for future data analysis and test results reporting activities. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to assure that they are reasonable and consistent

  2. Intercomparison of ozone measurements between Lidar and ECC-sondes

    Energy Technology Data Exchange (ETDEWEB)

    Grabbe, G.C. [Max-Planck-Institut fuer Meteorologie, Hamburg (Germany); Boesenberg, J. [Max-Planck-Institut fuer Meteorologie, Hamburg (Germany); Dier, H. [Meteorologisches Obs., Lindenberg (Germany); Goersdorf, U. [Meteorologisches Obs., Lindenberg (Germany); Matthias, V. [Max-Planck-Institut fuer Meteorologie, Hamburg (Germany); Peters, G. [Meteorologisches Obs., Lindenberg (Germany); Schaberl, T. [Hamburg Univ. (Germany). Meteorologisches Inst.; Senff, C. [Hamburg Univ. (Germany). Meteorologisches Inst.

    1996-02-01

    An intercomparison experiment for measurements of ozone vertical profiles in the lower troposphere was performed using a ground-based ozone DIAL (DIfferential Absorption Lidar) and ECC-sondes (Electrochemical Concentration Cell) attached to tethered as well as free flying balloons, which took place in June of 1994. The tethered balloon was used for ozone soundings in the planetary boundary layer up to an altitude of 500 m, while in the free troposphere free flying balloons were used. For the time series of up to 90 min obtained with the tethersondes both averages and standard deviations are compared. The mean difference for all measurements amounted to 3.5 {mu}g/m{sup 3} only, corresponding to 3.5%. For the instantaneous profiles compared to the free flying sondes the differences were only marginally larger, with a mean value of 3.6 {mu}g/m{sup 3} corresponding to 4.1%. With two exceptions all averages for a single profile stayed below 7 {mu}g/m{sup 3}. Larger individual excursions were observed. In some cases, in particular in regions of steep aerosol gradients at layer boundaries, most probably the lidar values cause the deviation, while in other cases the ECC-sonde is suspected to yield erroneous results. For the measured standard deviation those retrieved from DIAL measurements are generally larger than measured by the ECC-sondes, especially in regions of inhomogeneous aerosol distribution. For the measurements reported here, this is attributed to residual errors in the aerosol correction of the DIAL measurements. The general agreement found in this intercomparison is regarded as excellent, DIAL is proven to be a very valuable tool for detailed studies of the ozone distribution in the lower troposphere. (orig.)

  3. Using ECC DRAM to Adaptively Increase Memory Capacity

    OpenAIRE

    Luo, Yixin; Ghose, Saugata; Li, Tianshi; Govindan, Sriram; Sharma, Bikash; Kelly, Bryan; Boroumand, Amirali; Mutlu, Onur

    2017-01-01

    Modern DRAM modules are often equipped with hardware error correction capabilities, especially for DRAM deployed in large-scale data centers, as process technology scaling has increased the susceptibility of these devices to errors. To provide fast error detection and correction, error-correcting codes (ECC) are placed on an additional DRAM chip in a DRAM module. This additional chip expands the raw capacity of a DRAM module by 12.5%, but the applications are unable to use any of this extra c...

  4. Experiments on graphite block gaps connected with leak flow in bottom-core structure of experimental very high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Kikuchi, Kenji; Futakawa, Masatoshi; Takizuka, Takakazu; Kaburaki, Hideo; Sanokawa, Konomo

    1984-01-01

    In order to minimize the leak flow rate of an experimental VHTR (a multi-purpose very high-temperature gas-cooled reactor), the graphite blocks are tightened to reduce the gap distance between blocks by core restrainers surrounded outside of the fixed reflectors of the bottom-core structure and seal elements are placed in the gaps. By using a 1/2.75-scale model of the bottom-core structure, the experiments on the following items have been carried out: a relationship between core restraint force and block gap, a relationship between core restraint force and inclined angle of the model, leak flow characteristics of seal elements etc. The conclusions derived from the experiments are as follows: (1) Core restraint force is significantly effective for decreasing the gap distance between hot plenum blocks, but ineffective for the gap between hot plenum block and fixed reflector. (2) Graphite seal element reduces the leak flow rate from the top surface of hot plenum block into plenum region to one-third. (author)

  5. 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.

  6. Development and Modeling of Angled Effusion Cooling for the BR715 Low Emission Staged Combustor Core Demonstrator

    National Research Council Canada - National Science Library

    Gerendas, M

    2003-01-01

    .... The combustor cooling concept chosen was of the angled effusion type. Development of adequate modeling techniques and steady-state and transient rig tests to calibrate the thermal models was the key factor for the success...

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

    International Nuclear Information System (INIS)

    1982-09-01

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

  8. Disease Management of Early Childhood Caries: ECC Collaborative Project.

    Science.gov (United States)

    Ng, Man Wai; Ramos-Gomez, Francisco; Lieberman, Martin; Lee, Jessica Y; Scoville, Richard; Hannon, Cindy; Maramaldi, Peter

    2014-01-01

    Until recently, the standard of care for early childhood caries (ECC) has been primarily surgical and restorative treatment with little emphasis on preventing and managing the disease itself. It is now recognized that surgical treatment alone does not address the underlying etiology of the disease. Despite costly surgeries and reparative treatment, the onset and progression of caries are likely to continue. A successful rebalance of risk and protective factors may prevent, slow down, or even arrest dental caries and its progression. An 18-month risk-based chronic disease management (DM) approach to address ECC in preschool children was implemented as a quality improvement (QI) collaborative by seven teams of oral health care providers across the United States. In the aggregate, fewer DM children experienced new cavitation, pain, and referrals to the operating room (OR) for restorative treatment compared to baseline historical controls. The teams found that QI methods facilitated adoption of the DM approach and resulted in improved care to patients and better outcomes overall. Despite these successes, the wide scale adoption and spread of the DM approach may be limited unless health policy and payment reforms are enacted to compensate providers for implementing DM protocols in their practice.

  9. Disease Management of Early Childhood Caries: ECC Collaborative Project

    Directory of Open Access Journals (Sweden)

    Man Wai Ng

    2014-01-01

    Full Text Available Until recently, the standard of care for early childhood caries (ECC has been primarily surgical and restorative treatment with little emphasis on preventing and managing the disease itself. It is now recognized that surgical treatment alone does not address the underlying etiology of the disease. Despite costly surgeries and reparative treatment, the onset and progression of caries are likely to continue. A successful rebalance of risk and protective factors may prevent, slow down, or even arrest dental caries and its progression. An 18-month risk-based chronic disease management (DM approach to address ECC in preschool children was implemented as a quality improvement (QI collaborative by seven teams of oral health care providers across the United States. In the aggregate, fewer DM children experienced new cavitation, pain, and referrals to the operating room (OR for restorative treatment compared to baseline historical controls. The teams found that QI methods facilitated adoption of the DM approach and resulted in improved care to patients and better outcomes overall. Despite these successes, the wide scale adoption and spread of the DM approach may be limited unless health policy and payment reforms are enacted to compensate providers for implementing DM protocols in their practice.

  10. Considerations for Probabilistic Analyses to Assess Potential Changes to Large-Break LOCA Definition for ECCS Requirements

    International Nuclear Information System (INIS)

    Wilkowski, G.; Rudland, D.; Wolterman, R.; Krishnaswamy, P.; Scott, P.; Rahman, S.; Fairbanks, C.

    2002-01-01

    The U.S.NRC has undertaken a study to explore changes to the body of Part 50 of the U.S. Federal Code of Regulations, to incorporate risk-informed attributes. One of the regulations selected for this study is 10 CFR 50.46, A cceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors . These changes will potentially enhance safety and reduce unnecessary burden on utilities. Specific attention is being paid to redefining the maximum pipe break size for LB-LOCA by determining the spectrum of pipe diameter (or equivalent opening area) versus failure probabilities. In this regard, it is necessary to ensure that all contributors to probabilistic failures are accounted for when redefining ECCS requirements. This paper describes initial efforts being conducted for the U.S.NRC on redefining the LB-LOCA requirements. Consideration of the major contributors to probabilistic failure, and deterministic aspects for modeling them, are being addressed. At this time three major contributors to probabilistic failures are being considered. These include: (1) Analyses of the failure probability from cracking mechanisms that could involve rupture or large opening areas from either through-wall or surface flaws, whether the pipe system was approved for leak-before-break (LBB) or not. (2) Future degradation mechanisms, such as recent occurrence of PWSCC in PWR piping need to be included. This degradation mechanism was not recognized as being an issue when LBB was approved for many plants or when the initial risk-informed inspection plans were developed. (3) Other indirect causes of loss of pressure-boundary integrity than from cracks in the pipe system also should be included. The failure probability from probabilistic fracture mechanics will not account for these other indirect causes that could result in a large opening in the pressure boundary: i.e., failure of bolts on a steam generator manway, flanges, and valves; outside force damage from the

  11. CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Cleveland, J.C.

    1977-01-01

    CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

  12. 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.)

  13. Testing and performance analysis of a hollow fiber-based core for evaporative cooling and liquid desiccant dehumidification

    DEFF Research Database (Denmark)

    Jradi, Muhyiddine; Riffat, Saffa

    2016-01-01

    In this study, an innovative heat and mass transfer core is proposed to provide thermal comfort and humidity control using a hollow fiber contactor with multiple bundles of micro-porous hollow fibers. The hollow fiberbased core utilizes 12 bundles aligned vertically, each with 1,000 packed...

  14. Results of out-of-pile experiments to investigate the possibilities of cooling a core melt with internal heat production

    International Nuclear Information System (INIS)

    Fieg, G.

    1976-01-01

    After serious hypothetical reactor accidents, melted core materials with internal heat production can occur in large quantities. A retention of these molten core masses within the containment must be ensured. The knowledge of the heat transport from volume-heated layers is necessary to clarify this matter. (orig./LH) [de

  15. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon, E-mail: rjpark@kaeri.re.kr; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-03-15

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m{sup 2} s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  16. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-01-01

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m"2 s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  17. Melt cooling by bottom flooding: The experiment CometPC-H3. Ex-vessel core melt stabilization research

    International Nuclear Information System (INIS)

    Alsmeyer, H.; Cron, T.; Merkel, G.; Schmidt-Stiefel, S.; Tromm, W.; Wenz, T.

    2003-03-01

    The CometPC-H3 experiment was performed to investigate melt cooling by water addition to the bottom of the melt. The experiment was performed with a melt mass of 800 kg, 50% metal and 50% oxide, and 300 kW typical decay heat were simulated in the melt. As this was the first experiment after repair of the induction coil, attention was given to avoid overload of the induction coil and to keep the inductor voltage below critical values. Therefore, the height of the sacrificial concrete layer was reduced to 5 cm only, and the height of the porous concrete layers was also minimized to have a small distance and good coupling between heated melt and induction coil. After quite homogeneous erosion of the upper sacrificial concrete layer, passive bottom flooding started from the porous concrete after 220 s with 1.3 liter water/s. The melt was safely stopped, arrested and cooled. The porous, water filled concrete was only slightly attacked by the hot melt in the upper 25 mm of one sector of the coolant device. The peak cooling rate in the early contact phase of coolant water and melt was 4 MW/m 2 , and exceeded the decay heat by one order of magnitude. The cooling rate remarkably dropped, when the melt was covered by the penetrating water and a surface crust was formed. Volcanic eruptions from the melt during the solidification process were observed from 360 - 510 s and created a volcanic dome some 25 cm high, but had only minor effect on the generation of a porous structure, as the expelled melt solidified mostly with low porosity. Unfortunately, decay heat simulation in the melt was interrupted at 720 s by an incorrect safety signal, which excluded further investigation of the long term cooling processes. At that time, the melt was massively flooded by a layer of water, about 80 cm thick, and coolant water inflow was still 1 l/s. The melt had reached a stable situation: Downward erosion was stopped by the cooling process from the water filled, porous concrete layer. Top

  18. Combined group ECC protection and subgroup parity protection

    Science.gov (United States)

    Gara, Alan G.; Chen, Dong; Heidelberger, Philip; Ohmacht, Martin

    2013-06-18

    A method and system are disclosed for providing combined error code protection and subgroup parity protection for a given group of n bits. The method comprises the steps of identifying a number, m, of redundant bits for said error protection; and constructing a matrix P, wherein multiplying said given group of n bits with P produces m redundant error correction code (ECC) protection bits, and two columns of P provide parity protection for subgroups of said given group of n bits. In the preferred embodiment of the invention, the matrix P is constructed by generating permutations of m bit wide vectors with three or more, but an odd number of, elements with value one and the other elements with value zero; and assigning said vectors to rows of the matrix P.

  19. Numerical study of hot-leg ECC injection into the upper plenum of a pressurized water reactor

    International Nuclear Information System (INIS)

    Daly, B.J.; Torrey, M.D.; Rivard, W.C.

    1981-01-01

    In certain pressurized water reactor (PWR) designs, emergency core coolant (ECC) is injected through the hot legs into the upper plenum. The condensation of steam on this subcooled liquid stream reduces the pressure in the hot legs and upper plenum and thereby affects flow conditions throughout the reactor. In the present study, we examine countercurrent steam-water flow in the hot leg to determine the deceleration of the ECC flow that results from an adverse pressure gradient and from momentum exchange from the steam by interfacial drag and condensation. For the parameters examined in the study, water flow reversal is observed for a pressure drop of 22 to 32 mBar over the 1.5 m hot leg. We have also performed a three-dimensional study of subcooled water injection into air and steam environments of the upper plenum. The ECC water is deflected by an array of cylindrical guide tubes in its passage through the upper plenum. Comparisons of the air-water results with data obtained in a full scale experiment shows reasonable agreement, but indicates that there may be too much resistance to horizontal flow about the columns because of the use of a stair-step representation of the cylindrical guide tube cross section. Calculations of flow past single columns of stair-step, square and circular cross section do indicate excessive water deeentrainment by the noncircular column. This has prompted the use of an arbitrary mesh computational procedure to more accuratey represent the circular cross-section guide tubes. 15 figures

  20. Towards a spectroscopically accurate set of potentials for heavy hydride laser cooling candidates: Effective core potential calculations of BaH

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Keith; McLaughlin, Brendan M.; Lane, Ian C., E-mail: i.lane@qub.ac.uk [School of Chemistry and Chemical Engineering, Queen’s University Belfast, Stranmillis Road, Belfast BT9 5AG (United Kingdom)

    2016-04-14

    BaH (and its isotopomers) is an attractive molecular candidate for laser cooling to ultracold temperatures and a potential precursor for the production of ultracold gases of hydrogen and deuterium. The theoretical challenge is to simulate the laser cooling cycle as reliably as possible and this paper addresses the generation of a highly accurate ab initio {sup 2}Σ{sup +} potential for such studies. The performance of various basis sets within the multi-reference configuration-interaction (MRCI) approximation with the Davidson correction is tested and taken to the Complete Basis Set (CBS) limit. It is shown that the calculated molecular constants using a 46 electron effective core-potential and even-tempered augmented polarized core-valence basis sets (aug-pCVnZ-PP, n = 4 and 5) but only including three active electrons in the MRCI calculation are in excellent agreement with the available experimental values. The predicted dissociation energy D{sub e} for the X{sup 2}Σ{sup +} state (extrapolated to the CBS limit) is 16 895.12 cm{sup −1} (2.094 eV), which agrees within 0.1% of a revised experimental value of <16 910.6 cm{sup −1}, while the calculated r{sub e} is within 0.03 pm of the experimental result.

  1. The Merging Galaxy Cluster A520 - A Broken-Up Cool Core, A Dark Subcluster, and an X-Ray Channel

    Science.gov (United States)

    Wang, Qian H.S.; Markevitch, Maxim; Giacintucci, Simona

    2016-01-01

    We present results from a deep Chandra X-ray observation of a merging galaxy cluster A520. A high-resolution gas temperature map reveals a long trail of dense, cool clumpsapparently the fragments of a cool core that has been stripped from the infalling subcluster by ram pressure. The clumps should still be connected by the stretched magnetic field lines. The observed temperature variations imply that thermal conductivity is suppressed by a factor greater than 100 across the presumed direction of the magnetic field (as found in other clusters), and is also suppressed along the field lines by a factor of several. Two massive clumps in the periphery of A520, visible in the weak-lensing mass map and the X-ray image, have apparently been completely stripped of gas during the merger, but then re-accreted the surrounding high-entropy gas upon exit from the cluster. The mass clump that hosted the stripped cool core is also re-accreting hotter gas. An X-ray hydrostatic mass estimate for the clump that has the simplest geometry agrees with the lensing mass. Its current gas mass to total mass ratio is very low, 1.5 percent to 3 percent, which makes it a "dark subcluster." We also found a curious low X-ray brightness channel (likely a low-density sheet in projection) going across the cluster along the direction of an apparent secondary merger. The channel may be caused by plasma depletion in a region of an amplified magnetic field (with plasma Beta approximately equal to 10-20). The shock in A520 will be studied in a separate paper.

  2. AN X-RAY COOLING-CORE CLUSTER SURROUNDING A LOW-POWER COMPACT STEEP SPECTRUM RADIO SOURCE 1321+045

    International Nuclear Information System (INIS)

    Kunert-Bajraszewska, M.; Siemiginowska, A.; Labiano, A.

    2013-01-01

    We discovered an X-ray cluster in a Chandra observation of the compact steep spectrum (CSS) radio source 1321+045 (z = 0.263). CSS sources are thought to be young radio objects at the beginning of their evolution and can potentially test the cluster heating process. 1321+045 is a relatively low-luminosity source and its morphology consists of two radio lobes on the opposite sides of a radio core with no evidence for jets or hotspots. The optical emission line ratios are consistent with an interstellar medium dominated by active galactic nucleus photoionization with a small contribution from star formation, and no contributions from shocks. Based on these ratios, we classify 1321+045 as a low excitation galaxy (LEG) and suggest that its radioactivity is in a coasting phase. The X-ray emission associated with the radio source is detected with 36.1 ± 8.3 counts, but the origin of this emission is highly uncertain. The current X-ray image of the cluster does not show any signatures of a radio source impact on the cluster medium. Chandra detects the cluster emission at >3σ level out to ∼60'' (240 kpc). We obtain the best-fit beta model parameters of the surface brightness profile of β = 0.58 ± 0.2 and a core radius of 9.4 +1.1 -0.9 arcsec. The average temperature of the cluster is equal to kT = 4.4 +0.5 -0.3 keV, with a temperature and cooling profile indicative of a cooling core. We measure the cluster luminosity L (0.5-2 k eV) = 3 × 10 44 erg s –1 and mass 1.5 × 10 14 M ☉

  3. AN X-RAY COOLING-CORE CLUSTER SURROUNDING A LOW-POWER COMPACT STEEP SPECTRUM RADIO SOURCE 1321+045

    Energy Technology Data Exchange (ETDEWEB)

    Kunert-Bajraszewska, M. [Torun Centre for Astronomy, Faculty of Physics, Astronomy and Informatics, NCU, Grudziacka 5, 87-100 Torun (Poland); Siemiginowska, A. [Harvard Smithsonian Center for Astrophysics, 60 Garden St, Cambridge, MA 02138 (United States); Labiano, A., E-mail: magda@astro.uni.torun.pl [Centro de Astrobiologia (CSIC-INTA), Carretera de Ajalvir km. 4, E-28850 Torrejon de Ardoz, Madrid (Spain)

    2013-07-20

    We discovered an X-ray cluster in a Chandra observation of the compact steep spectrum (CSS) radio source 1321+045 (z = 0.263). CSS sources are thought to be young radio objects at the beginning of their evolution and can potentially test the cluster heating process. 1321+045 is a relatively low-luminosity source and its morphology consists of two radio lobes on the opposite sides of a radio core with no evidence for jets or hotspots. The optical emission line ratios are consistent with an interstellar medium dominated by active galactic nucleus photoionization with a small contribution from star formation, and no contributions from shocks. Based on these ratios, we classify 1321+045 as a low excitation galaxy (LEG) and suggest that its radioactivity is in a coasting phase. The X-ray emission associated with the radio source is detected with 36.1 {+-} 8.3 counts, but the origin of this emission is highly uncertain. The current X-ray image of the cluster does not show any signatures of a radio source impact on the cluster medium. Chandra detects the cluster emission at >3{sigma} level out to {approx}60'' (240 kpc). We obtain the best-fit beta model parameters of the surface brightness profile of {beta} = 0.58 {+-} 0.2 and a core radius of 9.4{sup +1.1}{sub -0.9} arcsec. The average temperature of the cluster is equal to kT = 4.4{sup +0.5}{sub -0.3} keV, with a temperature and cooling profile indicative of a cooling core. We measure the cluster luminosity L{sub (0.5-2{sub keV)}} = 3 Multiplication-Sign 10{sup 44} erg s{sup -1} and mass 1.5 Multiplication-Sign 10{sup 14} M{sub Sun}.

  4. 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

  5. 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.

  6. Finite element modeling of fluid/thermal/structural interaction for a gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    Bennett, J.G.; Ju, F.D.

    1980-01-01

    Two nonlinear finite element formulations for application to a series of experiments in the Gas-Cooled Fast Reactor (GCFR) development program are described. An efficient beam column element for moderately large deformations is combined with a finite element developed for an engineering description of a convecting fluid. Typical results from both elements are illustrated. A combined application for a problem typical of the GCFR loss-of-coolant experiments is illustrated. These problems are not the usual fluid structural interaction problems in that the inertia coupling is negligible while the thermal coupling is very important

  7. A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

    Science.gov (United States)

    Anghaie, S.; Chen, G.

    1996-01-01

    A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high

  8. 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

  9. Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Maschek, W.; Royl, P.

    1988-09-01

    Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

  10. Full Core Criticality Modeling of Gas-Cooled Fast Reactor Using the SCALE6.0 and MCNP5 Code Packages

    International Nuclear Information System (INIS)

    Matijevic, M.; Jecmenica, R.; Pevec, D.; Trontl, K.

    2012-01-01

    The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals of the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working enviroment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 code system, with the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analysed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be

  11. Hypothetical core disruptive accident analysis of a 2000 MWsub(e) liquid metal cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Struwe, D.

    1977-12-01

    A structural phase diagram for hypothetical core disruptive accidents (HCDA) has been developed based on a variety of analyses for different LMFBR's. The intention was to identify the strategic phases of HCDA's important with regard to safety aspects of the plant. These phases are investigated in detail for a 2,000 MWsub(e) LMFBR (SNR-2,000). Characteristic data of SNR-2,000 are discussed concerning their influence on safety analysis. Reasons for the choice of model parameters for special phenomena as fuel coolant interaction, fuel pin failure mechanisms and sodium voiding are given. The results of calculations with CAPRI-2, HOPE and KADIS are analyzed for possibilities to enter energetic core disassembly with consequences, making power values below 2,000 MWsub(e) necessary. Investigation of these results shows that the expected consequences do not lead to design requirements, restricting the magnitude of the electrical power output of LMFBR's to values below 2,000 MWsub(e). Therefore, consequences of HCDA's are principal not expected to limit the feasibility of conventional core design of this order of magnitude. (orig.) [de

  12. Estimated net value and uncertainty for automating ECCS switchover at PWRs

    International Nuclear Information System (INIS)

    Walsh, B.; Brideau, J.; Comes, L.; Darby, J.; Guttmann, H.; Sciacca, F.; Souto, F.; Thomas, W.; Zigler, G.

    1996-02-01

    Question for resolution of Generic Safety Issue No. 24 is whether or not PWRs that currently rely on a manual system for ECCS switchover to recirculation should be required to install an automatic system. Risk estimates are obtained by reevaluating the contributions to core damage frequencies (CDFs) associated with failures of manual and semiautomatic switchover at a representative PWR. This study considers each separate instruction of the corresponding emergency operating procedures (EOPs), the mechanism for each control, and the relation of each control to its neighbors. Important contributions to CDF include human errors that result in completely coupled failure of both trains and failure to enter the required EOP. It is found that changeover to a semiautomatic system is not justified on the basis of cost-benefit analysis: going from a manual to a semiautomatic system reduces the CDF by 1.7 x 10 -5 per reactor-year, but the probability that the net cost of the modification being less than $1, 000 per person-rem is about 20% without license renewal. Scoping analyses, using optimist assumptions, were performed for a changeover to a semiautomatic system with automatic actuation and to a fully automatic system; in these cases the probability of a net cost being less than $1,000/person-rem is about 50% without license renewal and over 95% with license renewal

  13. Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Okubo, Tsutomu

    2009-01-01

    The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.

  14. Magnitude and reactivity consequences of accidental moisture ingress into the Modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Smith, O.L.

    1992-01-01

    Accidental admission of moisture into the primary system of a Modular High-Temperature Gas-Cooled Reactor (MHTGR) has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moistureingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. Rate and magnitude of steam buildup are found to be dominated by major system features such as break size in comparison with safety valve capacity and reliability, while being less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  15. Magnitude and reactivity consequences of moisture ingress into the modular High-Temperature Gas-Cooled Reactor core

    International Nuclear Information System (INIS)

    Smith, O.L.

    1992-12-01

    Inadvertent admission of moisture into the primary system of a modular high-temperature gas-cooled reactor has been identified in US Department of Energy-sponsored studies as an important safety concern. The work described here develops an analytical methodology to quantify the pressure and reactivity consequences of steam-generator tube rupture and other moisture-ingress-related incidents. Important neutronic and thermohydraulic processes are coupled with reactivity feedback and safety and control system responses. The rate and magnitude of steam buildup are found to be dominated by major system features such as break size compared with safety valve capacity and reliability and less sensitive to factors such as heat transfer coefficients. The results indicate that ingress transients progress at a slower pace than previously predicted by bounding analyses, with milder power overshoots and more time for operator or automatic corrective actions

  16. Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

    1980-11-01

    As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

  17. 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

  18. Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

    Science.gov (United States)

    Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

    2018-01-01

    ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical

  19. Development of C/C composite for the core component of the high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H

    2005-01-15

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite.

  20. Development of C/C composite for the core component of the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Park, J. Y.; Kim, W. J.; Ryu, W. S.; Jang, J. H.

    2005-01-01

    This report reviewed a state of the art on development of C/C composite for the core components for VHTR and described the followings items. The fabrication methods of C/C composites. Summary on the JAERI report (JAERI-Res 2002-026) on the process screening test for the selection of a proper C/C composite material. Review of the proceedings presented at the GEN-IV VHTR material PMB meeting. A status of the domestic commercial C/C composite. The published property data and the characteristics of the commercial C/C composite

  1. Investigating the Role of Shell Thickness and Field Cooling on Saturation Magnetization and Its Temperature Dependence in Fe3O4/γ-Fe2O3 Core/Shell Nanoparticles

    Directory of Open Access Journals (Sweden)

    Ihab M. Obaidat

    2017-12-01

    Full Text Available Understanding saturation magnetization and its behavior with particle size and temperature are essential for medical applications such magnetic hyperthermia. We report the effect of shell thickness and field cooling on the saturation magnetization and its behavior with temperature in Fe3O4/γ-Fe2O3 core/shell nanoparticles of fixed core diameter (8 nm and several shell thicknesses. X-ray diffraction (XRD analysis and transmission electron microscopy (TEM, high-resolution transmission electron microscopy (HRTEM were used to investigate the phase and the morphology of the samples. Selected area electron diffraction (SAED confirmed the core/shell structure and phases. Using a SQUID (San Diego, CA, USA, magnetic measurements were conducted in the temperature range of 2 to 300 K both under zero field-cooling (ZFC and field-cooling (FC protocols at several field-cooling values. In the ZFC state, considerable enhancement of saturation magnetization was obtained with the increase of shell thickness. After field cooling, we observed a drastic enhancement of the saturation magnetization in one sample up to 120 emu/g (50% larger than the bulk value. In both the FC and ZFC states, considerable deviations from the original Bloch’s law were observed. These results are discussed and attributed to the existence of interface spin-glass clusters which are modified by the changes in the shell thickness and the field-cooling.

  2. Predictive Eco-Cruise Control (ECC) system : model development, modeling and potential benefits.

    Science.gov (United States)

    2013-02-01

    The research develops a reference model of a predictive eco-cruise control (ECC) system that intelligently modulates vehicle speed within a pre-set speed range to minimize vehicle fuel consumption levels using roadway topographic information. The stu...

  3. Application of Interfacial Propagation and Kinking Crack Concept to ECC/Concrete Overlay Repair System

    Directory of Open Access Journals (Sweden)

    Yaw ChiaHwan

    2014-01-01

    Full Text Available Research on the application of ultraductile engineered cementitious composite (ECC as overlay in the repair of deteriorated concrete structures is performed in this paper. Also, interfacial crack kinking and trapping mechanism experimentally observed in ECC/concrete overlay repair system are described by comparison of toughness and energy release rate. The mechanism involves cycles of extension, kinking, and arrest of interfacial crack into the overlay. Experimental testing of overlay repair system reveals significant improvements in load carrying capacity and ductility over conventional concrete overlay. The commonly observed overlay system failure mode of delamination or spalling is eliminated when ECC is applied. These failure modes are suppressed when ECC is used as an ideal and durable candidate overlay repair material.

  4. Preestrita pihtimus : Friedrich Nietzsche ja tema substantsiaalne mina (Ecce homo) / Jaan Undusk

    Index Scriptorium Estoniae

    Undusk, Jaan, 1958-

    1998-01-01

    Varem ilmunud raamatu järelsõnana: Nietzsche, Friedrich. Ecce homo : kuidas saadakse selleks, mis ollakse / tlk. Jaan Undusk. Tallinn : Vagabund, 1996. Sisu: Poeetiline, filosoofiline, dionüüsiline kaanon ; Künism contra hulluskahtlus ; Antikristlik pihtimus ; Egoretoorika

  5. Reference core design Mark-III of the experimental multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Watanabe, Takashi; Ishiguro, Okikazu; Kuroki, Syuzi

    1977-10-01

    The reactivity control system is one of the important items in reactor design, but it is much restricted by structural design of fuel element and pressure vessel in the experimental multi-purpose, high-temperature reactor. Preceding the first conceptual design of the reactor, therefore, the reactivity control system composed of control rod, burnable poison and reserve shutdown system in Mark-II design was re-studied, and several improvements were indicated. (1) The diameter of control rods must be as large as possible because it is impossible to increase the number of control rods. (2) The accuracy in estimation of the reactivity to be compensated with control rods is important because of the mutual interference of pair control rods with the twin configuration in a fuel element. (3) The improvement of core performance in burnup is accompanied by the reduction of design margin for control rods. (4) Increase of the reactivity to be compensated with the burnable poison leads to increase of the core reactivity recovery with burnup, and the assertion of the decrease for recovery of reactivity leads to increase of the temperature dependency of reactivity compensated with control rods. (5) Reduction of reactivity to be compensated with control rods is thus limited by cancellation of the effects in the reactivity recovery and the reactivity temperature dependency. (6) The reserve shutdown system can be designed with margin under the condition of excluding the reactivity of burnup from that to be compensated. (auth.)

  6. Estimative of core damage frequency in IPEN's IEA-R1 research reactor (PSA level 1) due to the initiating event of loss of coolant caused by large rupture in the pipe of the primary circuit

    International Nuclear Information System (INIS)

    Hirata, Daniel Massami

    2009-01-01

    This work applies the methodology of probabilistic safety assessment level 1 to the research reactor IEA-R1 IPEN-CNEN/SP. Two categories of identified initiating events of accidents in the reactor are studied: loss of flow and loss of primary coolant. Among the initiating events, blockage of flow channel and loss of cooling fluid by major pipe rupture in the primary circuit are chosen for a detailed analysis. The event tree technique is used to analyze the evolution of the accident, including the actuation or the fail of actuation of the safety systems and the reactor damages. Using the fault tree the reliability of the following reactor safety systems is evaluated: reactor shutdown system, isolation of the reactor pool, emergency core cooling system (ECCS) and the electric system. Estimative for the frequency of damage to the reactor core and the probability of failure of the analyzed systems are calculated. The estimated values for the frequencies of core damage are within the expected margins and are of the same order of magnitude as those found for similar reactors. The reliability of the reactor shutdown system, isolation of the reactor pool and ECCS are satisfactory for the conditions these systems are required. However, for the electric system it is suggested an upgrade to increase its reliability. (author)

  7. An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, M.; Farmer, M.; Grabaskas, D.

    2017-06-26

    The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

  8. 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.)

  9. Stress relaxation and creep of high-temperature gas-cooled reactor core support ceramic materials: a literature search

    International Nuclear Information System (INIS)

    Selle, J.E.; Tennery, V.J.

    1980-05-01

    Creep and stress relaxation in structural ceramics are important properties to the high-temperature design and safety analysis of the core support structure of the HTGR. The ability of the support structure to function for the lifetime of the reactor is directly related to the allowable creep strain and the ability of the structure to withstand thermal transients. The thermal-mechanical response of the core support pads to steady-state stresses and potential thermal transients depends on variables, including the ability of the ceramics to undergo some stress relaxation in relatively short times. Creep and stress relaxation phenomena in structural ceramics of interest were examined. Of the materials considered (fused silica, alumina, silicon nitride, and silicon carbide), alumina has been more extensively investigated in creep. Activation energies reported varied between 482 and 837 kJ/mole, and consequently, variations in the assigned mechanisms were noted. Nabarro-Herring creep is considered as the primary creep mechanism and no definite grain size dependence has been identified. Results for silicon nitride are in better agreement with reported activation energies. No creep data were found for fused silica or silicon carbide and no stress relaxation data were found for any of the candidate materials. While creep and stress relaxation are similar and it is theoretically possible to derive the value of one property when the other is known, no explicit demonstrated relationship exists between the two. For a given structural ceramic material, both properties must be experimentally determined to obtain the information necessary for use in high-temperature design and safety analyses

  10. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall

  11. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

  12. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    International Nuclear Information System (INIS)

    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

  13. An ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Anish, E-mail: anish@igcar.gov.in; Rajkumar, K.V.; Sharma, Govind K.; Dhayalan, R.; Jayakumar, T.

    2015-02-15

    Highlights: • We demonstrate a novel ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast breeder reactor. • The methodology comprises of the inspection of shell weld immersed in sodium from the outside surface of the main vessel using ultrasonic guided wave. • The formation and propagation of guided wave modes are validated by finite element simulation of the inspection methodology. • A defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably using the developed methodology. - Abstract: The paper presents a novel ultrasonic methodology developed for in-service inspection (ISI) of shell weld of core support structure of main vessel of 500 MWe prototype fast breeder reactor (PFBR). The methodology comprises of the inspection of shell weld immersed in sodium from the outsider surface of the main vessel using a normal beam longitudinal wave ultrasonic transducer. Because of the presence of curvature in the knuckle region of the main vessel, the normal beam longitudinal wave enters the support shell plate at an angle and forms the guided waves by mode conversion and multiple reflections from the boundaries of the shell plate. Hence, this methodology can be used to detect defects in the shell weld of the core support structure. The successful demonstration of the methodology on a mock-up sector made of stainless steel indicated that an artificial defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably.

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

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, LianCheng; Zhang, Bin

    2016-01-01

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

  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. 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. 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

  18. In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

    In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

  19. Reliability investigation for the ECC subsystem of a 1300 MWe-PWR

    International Nuclear Information System (INIS)

    Lalovic, M.

    1983-01-01

    In this study, a fault-tree analysis is used for reliability investigation of Emergency Core Cooling Sub-system of a 1300 MWe pressurised water reactor. Basic assumptions of the study are large break in the reactor coolant system and independence of the pseudo-components. Relatively high non-availability of the sub-system was calculated. Critical component and minimum cut set are determined. (author)

  20. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Maldonado, Ivan [Univ. of Tennessee, Knoxville, TN (United States)

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  1. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Maldonado, Ivan

    2016-01-01

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate ('plank') fuel. Proposal to FY12 NEUP entitled 'Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors' was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed on December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project's success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.

  2. The prevalence and risks of early childhood caries (ECC) in Toronto, Canada.

    Science.gov (United States)

    Al-Jewair, Thikriat S; Leake, James L

    2010-10-14

    To determine the prevalence and risks of early childhood caries (ECC) among children less than 71 months of age in Toronto, Canada, and to evaluate the association between parental/caregiver depression and ECC. A secondary analysis of data previously collected by the Toronto Public Health as part of the 2003 Toronto Perinatal and Child Health Survey was performed. The 90-item survey was conducted over the telephone to 1,000 families with children from zero years (birth) to six years of age. Parents/caregivers were asked about factors related to the development and health of their children. For this study, only children younger than six years of age (less than 71 months) were included (n=833). The primary outcome of interest was self-reported and measured by the response to the question of whether a physician/dentist had ever told the parent/caregiver his/her child had ECC. The prevalence of ECC was 4.7 percent (37 of 791 children). The child's age, his/her history of dental visits, teeth brushing, the use of fluoridated toothpaste, the parent's/caregiver's depressive tendencies, the language spoken at home, and the household annual income were all significant in the bivariate analysis. Multiple logistic regression identified four factors associated with ECC: the child's age (being three years of age or older), having at least one parent/caregiver with depression, not speaking English at home, and having an annual household income less than $40,000 in Canadian dollars (CAD). While a child's age, home language, and household income are known risks for ECC, the finding that parental/caregiver depression may be related to ECC is new. Multiple risk factors are involved in the development of early childhood caries. Of particular importance are demographic (e.g., child's age), social (e.g., annual household income), and psychosocial factors (e.g., parental/caregiver depression) that are indirectly linked to ECC. More attention needs to be placed on understanding the role

  3. Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

    International Nuclear Information System (INIS)

    Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

    2010-01-01

    To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

  4. Direct ECC bypass phenomena in the MIDAS test facility during LBLOCA reflood phase

    International Nuclear Information System (INIS)

    Yun, B. J.; Kweon, T. S.; Ah, D. J.; Ju, I. C.; Song, C. H.; Park, J. K.

    2001-01-01

    This paper describes the experimental results of ECC Direct Bypass Phenomena in the downcomer during the late reflood phase of LBLOCA of the reactor that adopts Direct Vessel Injection as a ECC system. The experiments have been performed in MIDAS test facility using superheated steam and water. The test condition was determined, based on the preliminary analysis of TRAC code, from modified linear scaling method of 1/4.93 length scale. To measure the direct bypass fraction according to the nozzle location, separate effect tests have been performed in case of DVI-4(farthest from broken cold leg) injection, DVI-2(closest to broken cold leg) injection, and DVI-2 and 4 injection, respectively. Also the test was carried out varying the steam flow rate greatly to investigate the effect of steam flow rate on the direct bypass fraction of ECC water. Test results show that the direct bypass fraction of ECC water depends significantly on the injected steam mass flow rate. DVI-4 tests show that the direct bypass fraction increases drastically as the steam flow rate increases. However, in DVI-2 test most of the injected ECC water penetrates into lower downcomer. The direct bypass characteristic in each of DVI-2 and DVI-4 tests is reflected into the direct bypass characteristic curve of DVI-2 and 4 test. The steam condensation reaches to the theoretically allowable maximum value

  5. Energy-dissipating and self-repairing SMA-ECC composite material system

    International Nuclear Information System (INIS)

    Li, Xiaopeng; Li, Mo; Song, Gangbing

    2015-01-01

    Structural component ductility and energy dissipation capacity are crucial factors for achieving reinforced concrete structures more resistant to dynamic loading such as earthquakes. Furthermore, limiting post-event residual damage and deformation allows for immediate re-operation or minimal repairs. These desirable characteristics for structural ‘resilience’, however, present significant challenges due to the brittle nature of concrete, its deformation incompatibility with ductile steel, and the plastic yielding of steel reinforcement. Here, we developed a new composite material system that integrates the unique ductile feature of engineered cementitious composites (ECC) with superelastic shape memory alloy (SMA). In contrast to steel reinforced concrete (RC) and SMA reinforced concrete (SMA-RC), the SMA-ECC beams studied in this research exhibited extraordinary energy dissipation capacity, minimal residual deformation, and full self-recovery of damage under cyclic flexural loading. We found that the tensile strain capacity of ECC, tailored up to 5.5% in this study, allows it to work compatibly with superelastic SMA. Furthermore, the distributed microcracking damage mechanism in ECC is critical for sufficient and reliable recovery of damage upon unloading. This research demonstrates the potential of SMA-ECC for improving resilience of concrete structures under extreme hazard events. (paper)

  6. Holistic care patient with Early Childhood Caries (ECC: A case report

    Directory of Open Access Journals (Sweden)

    Intan Maulani

    2016-06-01

    Full Text Available Early Childhood Caries (ECC is a specific form of severe dental caries that affects infants and young children. ECC progresses rapidly in those who are at high risk, and often goes untreated. Children experiencing caries as infants or toddlers have a much greater probability of subsequent caries in both primary and permanent dentitions. This case showed management holistic care for children with ECC.A five year old boy patient accompanied with her parents were reported to the Pedodontic Clinic Padjadjaran University Dental Hospital with a chief complaint of decayed upper anterior teeth and pain in the molar teeth. Clinical examinations found dental caries almost all teeth in the maxilla and mandible. Based on panoramic radiograph, treatments that can be done are strip crown glass ionomer restorations, pulp treatments, extractions and fixed space maintainer. Patients diagnosed with severe ECC, patient and parents described on this type of caries. During treatment the patient was given oral hygiene instruction and recommend daily use of tooth mouse. After all treatment were completed, fluoride topical, and fissure sealants, recall check up after three months was scheduled. Holistic care needed in handling children with ECC.

  7. Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

    1985-04-01

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

  8. Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Esquivel E, J.

    2016-09-01

    The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

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

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu; Ohkubo, Yoshiyuki

    2010-01-01

    Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot under transient conditions from forced to natural circulation operations that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal, e.g., flow redistribution in fuel assemblies caused by buoyancy force, and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional 1D and detailed 3D simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness. (author)

  10. Design and safety aspect of lead and lead-bismuth cooled long-life small safe fast reactors for various core configurations

    International Nuclear Information System (INIS)

    Zaki, S.; Sekimoto, Hiroshi

    1995-01-01

    Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ∼2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25-40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient. (author)

  11. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    International Nuclear Information System (INIS)

    Basol, Mit; Kielb, John F.; MuHooly, John F.; Smit, Kobus

    2007-01-01

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  12. On the energy shift the ECC cusp. Does the shift really exist?

    International Nuclear Information System (INIS)

    Sarkadi, L.; Barrachina, R.O.

    2004-01-01

    The cusplike 'electron capture to the continuum' (ECC) peak appearing in the spectrum of the forward emitted electrons in ion-atom collisions are generally thought to be a divergence. Recently Shah et al., however, claimed that 'the ECC cusp is indeed a cusp, and not a divergence smoothed by the experiment'. These authors measured the ECC cusp for collisions of 10- and 20-keV protons with H 2 and He, and found that the peak was shifted to lower velocity than its expected position. We also carried out CTMC calculations (for the case of 20-keV protons on He) by which we demonstrated that the shift really exists but its value depends on the angular window of the electron detection. (K.A.)

  13. Lightweight ECC based RFID authentication integrated with an ID verifier transfer protocol.

    Science.gov (United States)

    He, Debiao; Kumar, Neeraj; Chilamkurti, Naveen; Lee, Jong-Hyouk

    2014-10-01

    The radio frequency identification (RFID) technology has been widely adopted and being deployed as a dominant identification technology in a health care domain such as medical information authentication, patient tracking, blood transfusion medicine, etc. With more and more stringent security and privacy requirements to RFID based authentication schemes, elliptic curve cryptography (ECC) based RFID authentication schemes have been proposed to meet the requirements. However, many recently published ECC based RFID authentication schemes have serious security weaknesses. In this paper, we propose a new ECC based RFID authentication integrated with an ID verifier transfer protocol that overcomes the weaknesses of the existing schemes. A comprehensive security analysis has been conducted to show strong security properties that are provided from the proposed authentication scheme. Moreover, the performance of the proposed authentication scheme is analyzed in terms of computational cost, communicational cost, and storage requirement.

  14. Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    feedback (Doppler and displacement) VENUS-II has the following assumptions. The reactor materials behave like a homogeneous mixture with the property of an isotropic and nonviscous fluid. The reactivity change caused by a material displacement can be calculated with first-order perturbation theory. Further, the reactivity worth of spatial gradients remain constant and distort with the grid. The heat transfer from the fuel can be ignored. Although several heat transfer mechanisms can become significant, one of the greatest potential influence would appear to be a rapid molten-fuel-coolant interaction (MFCI). The non fuel core constituents are considered to be compressible, but inert, materials. The fuel vapor pressure and compression of the reactor materials are the only sources of internal pressure. Thus, such potential pressure sources such as fission gas and sodium vapor pressure are ignored. The time history of the power level can be described using point kinetics, and the spatial power-density distribution remains constant. In this work, the energy released from core disruptive accident (CDA) of sodium cooled fast reactor was investigated using CDA-ER and VENUS-II code for various reactivity insertion rates up to 100$/s, and their results were compared. The calculation results of two codes showed similar trends of energy, power and pressure from CDA. But most results of VENUS-II were found to be larger than those of CDA-ER. The released energy results calculated from VENUS-II were about 2 ∼ 3 times higher than those from CDA-ER.

  15. Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

    International Nuclear Information System (INIS)

    Kang, S. H.; Ha, K. S.

    2013-01-01

    displacement) VENUS-II has the following assumptions. The reactor materials behave like a homogeneous mixture with the property of an isotropic and nonviscous fluid. The reactivity change caused by a material displacement can be calculated with first-order perturbation theory. Further, the reactivity worth of spatial gradients remain constant and distort with the grid. The heat transfer from the fuel can be ignored. Although several heat transfer mechanisms can become significant, one of the greatest potential influence would appear to be a rapid molten-fuel-coolant interaction (MFCI). The non fuel core constituents are considered to be compressible, but inert, materials. The fuel vapor pressure and compression of the reactor materials are the only sources of internal pressure. Thus, such potential pressure sources such as fission gas and sodium vapor pressure are ignored. The time history of the power level can be described using point kinetics, and the spatial power-density distribution remains constant. In this work, the energy released from core disruptive accident (CDA) of sodium cooled fast reactor was investigated using CDA-ER and VENUS-II code for various reactivity insertion rates up to 100$/s, and their results were compared. The calculation results of two codes showed similar trends of energy, power and pressure from CDA. But most results of VENUS-II were found to be larger than those of CDA-ER. The released energy results calculated from VENUS-II were about 2 ∼ 3 times higher than those from CDA-ER

  16. BREASTFEEDING AND EARLY CHILDHOOD CARIES (ECC SEVERITY OF CHILDREN UNDER THREE YEARS OLD IN DKI JAKARTA

    Directory of Open Access Journals (Sweden)

    Risqa Rina Darwita

    2008-12-01

    Full Text Available The prevalence and severity of caries in children under three years old are constantly increasing. One of the cause is the increase consumption of cariogenic carbohydrate. Breast milk have buffer capacity that eventually able to prevent caries. The aims of this research are to discover the correlation between breastfeeding with the severity of Early Childhood Caries (ECC in children under three years old, to provide information on prevalence and severity of caries in children under three years, and to explain factors influencing the incidence of ECC. This research designed cross sectionally and conducted upon 565 children aged 12-38 months, selected using multistage cluster random sampling. The ECC prevalence of children under three years in DKI Jakarta is 52.7%, with average score of def-t 2.85. Bivariate analysis showed that, variables which correlate with the level of ECC are; the way to deliver (p=0,012, frequency (p=0,002, duration (p=0,002, salivary buffer capacity (p=0.013, habitual consumption of sugary diet (p=0.005, child’s dental hygiene behavior (p=0.002, and mother’s education (p=0.001. Multivariate analysis showed that ECC can be explained by these variables: age, mother’s education, the way to deliver and frequency of breast milk complements/replacement of consumtions, child’s brushing habit, plaque pH, and salivary buffer capacity. Resulting determination coefficient 32.1%. There is no significant correlation between breastfeeding and the level of Early Childhood Caries (ECC. The role of protective qualities of breastfeeding are not shown because of bias in obtaining data influence the incidence of caries.

  17. Gas cooled reactors

    International Nuclear Information System (INIS)

    Kojima, Masayuki.

    1985-01-01

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

  18. Is 300 Seconds ACT Safe and Efficient during MiECC Procedures?

    Science.gov (United States)

    Bauer, Adrian; Hausmann, Harald; Schaarschmidt, Jan; Szlapka, Michal; Scharpenberg, Martin; Eberle, Thomas; Hasenkam, J Michael

    2017-12-31

     The recommended minimum activated clotting time (ACT) level for cardiopulmonary bypass (CPB) of 480 seconds originated from investigations with bubble oxygenators and uncoated extracorporeal circulation (ECC) systems. Modern minimal invasive ECC (MiECC) systems are completely closed circuits containing a membrane oxygenator and a tip-to-tip surface coating. We hypothesized that surface coating and the "closed-loop" design allow the MiECC to safely run with lower ACT levels and that an ACT level of 300 seconds can be safely applied without thromboembolic complications. The aim of this study was to investigate the potential risks during application of reduced heparin levels in patients undergoing coronary surgery.  In this study, 68 patients undergoing coronary artery bypass grafting with MiECC were randomized to either the study group with an ACT target of 300 seconds or the control group with an ACT of 450 seconds. All other factors of MiECC remained unchanged.  The study group received significantly less heparin and protamine (heparin [international units] median [min-max], Red_AC: 32,800 [23,000-51,500] vs. Full_AC: 50,000 [35,000-65,000] p  ACT in the study group was significantly lower at the start of MiECC (mean ± standard deviation: study group 400 ± 112 vs. control group 633 ± 177; p  ACT levels were: study group 344 ± 60 versus control group 506 ± 80. In both groups, the values of the endogenous thrombin potential (ETP) decreased simultaneously. None of the study participants experienced thromboembolic complications.  Since no evidence of increased thrombin formation (ETP) was found from a laboratory standpoint, we concluded that the use of MiECC with a reduced anticoagulation strategy seems possible. This alternative anticoagulation strategy leads to significant reduction in dosages of both heparin and protamine. We can confidently move forward with investigating this anticoagulation concept. However, to

  19. ECC delivery to lower plenum under downcomer injection part 2. RELAP5 assessment

    International Nuclear Information System (INIS)

    Bang, Young Seok; Shin, An Dong; Kim, Hho Jung

    2000-01-01

    In the present study, the capability of the thermal-hydraulic codes, RELAP5/MOD3.2.2 gamma, in predicting the steam-water interaction and the related ECC delivery to lower plenum under downcomer injection condition during refill phase is evaluated using the experimental data of the UPTF Test 21A. The facility is modeled in detail, and the test condition simulated for code calculations. The calculation result is compared with the applicable measurement data and discussed for the pressure response, ECC bypass behavior, lower plenum delivery, global water mass distribution, and local behavior in downcomer

  20. Preservation of Postharvest Quality of Leafy Amaranth (Amaranthus spp. Vegetables Using Evaporative Cooling

    Directory of Open Access Journals (Sweden)

    Jane Ambuko

    2017-01-01

    Full Text Available Leafy vegetables are very highly perishable and must be utilized immediately after harvest. Their fast deterioration is attributed to various biological and environmental factors with temperature playing a central role. Evaporative cooling is a low-cost temporary storage technology that offers smallholder vegetable farmers an alternative to expensive cold rooms. The present study sought to determine the effectiveness of evaporative cooling using zero energy brick cooler (ZEBC and evaporative charcoal cooler (ECC, to preserve the postharvest quality of leafy amaranth vegetables. Freshly harvested vegetables were separated into bundles weighing 300 grams and stored under ZEBC, ECC, and ambient room conditions (control. Real time changes in temperature and relative humidity (RH as well as changes in quality attributes (physiological weight loss (PWL, wilting index, hue angle, and vitamin C were determined during the storage period. The temperature difference between the ZEBC and ECC versus the ambient air ranged between 4 and 10°C. Significantly higher RH (80–100% was recorded in both evaporative cooling chambers. At the end of storage, higher PWL (47.6% was recorded at ambient room conditions compared to 10.5 and 6.7% under ZEBC and ECC, respectively. A rapid decline in vitamin C (51% was reported in vegetables stored at ambient room conditions. Overall, there was better vegetable quality preservation under ECC and ZEBC.

  1. Passive residual energy utilization system in thermal cycles on water-cooled power reactors

    International Nuclear Information System (INIS)

    Placco, Guilherme M.; Guimaraes, Lamartine N.F.; Santos, Rubens S. dos

    2013-01-01

    This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

  2. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

    A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

  3. ROSA-V/LSTF vessel top head LOCA tests SB-PV-07 and SB-PV-08 with break sizes of 1.0 and 0.1% and operator recovery actions for core cooling

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2010-02-01

    A series of break size parameter tests (SB-PV-07 and SB-PV-08) were conducted at the Large Scale Test Facility (LSTF) of ROSA-V Program by simulating a vessel top small break loss-of-coolant accident (SBLOCA) at a pressurized water reactor (PWR). Typical phenomena to the vessel top break LOCA and effectiveness of operator recovery actions on core cooling were studied under an assumption of total failure of high pressure injection (HPI) system. The LSTF simulates a 4-loop 3423 MWt PWR by a full-height, full-pressure and 1/48 volume scaling two-loop system. Typical phenomena of vessel top break LOCA are clarified for the cases with break sizes of 1.0 and 0.1% cold leg break equivalent. The results from a 0.5% top break LOCA test (SB-PV-02) in the early ROSA-IV Program was referred during discussion. Operator actions of HPI recovery in the 1.0% top break test and steam generator (SG) depressurization in the 0.1% top break test were initiated when temperature at core exit thermocouple (CET) reached 623 K during core boil-off. Both operator actions resulted in immediate recovery of core cooling. Based on the obtained data, several thermal-hydraulic phenomena were discussed further such as relations between vessel top head water level and steam discharge at the break, and between coolant mass inventory transient and core heat-up and quench behavior, and CET performances to detect core heat-up under influences of three-dimensional (3D) steam flows in the core and core exit. (author)

  4. Cold leg injection reflood test results in the SCTF Core-I under constant system pressure

    International Nuclear Information System (INIS)

    Adachi, Hiromichi; Iwamura, Takamichi; Sobajima, Makoto; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio.

    1990-08-01

    The Slab Core Test Facility (SCTF) was constructed to investigate two-dimensional thermal-hydrodynamics in the core and the interaction in fluid behavior between the core and the upper plenum during the last part of blowdown, refill and reflood phases of a postulated loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The present report describes the analytical results on the system behavior observed in the SCTF Core-I cold leg injection tests, S1-14 (Run 520), S1-15 (521), S1-16 (522), S1-17 (523), S1-20 (530), S1-21 (531), S1-23 (536) and S1-24 (537), performed under constant system pressure condition during transient. Major discussion items are: (1) steam binding, (2) U-tube oscillations, (3) bypass of ECC water (4) core cooling behavior, (5) effect of vent valve and (6) other parameter effects. These results give us very useful information and suggestion on reflood behavior. (author)

  5. MICROORGANISMS FOUND IN SECONDARY CATARACT MATERIAL OF ECCE PATIENTS, A STUDY WITH SEM AND TEM

    NARCIS (Netherlands)

    KALICHARAN, D; JONGEBLOED, WL; LOS, LI; WORST, JGF

    1993-01-01

    Globular secondary cataract material, removed from 24 patients with ECCE after ophthalmic cleaning of the anterior capsule, were investigated with SEM and TEM. Besides spherical, somewhat oval shaped bodies of various shape and size comparable with those found in cataractous lenses, (an)aerobic

  6. Complete genome sequences of Escherichia coli strains 1303 and ECC-1470 isolated from bovine mastitis

    NARCIS (Netherlands)

    Leimbach, Andreas; Poehlein, Anja; Witten, Anika; Scheutz, Flemming; Schukken, Ynte|info:eu-repo/dai/nl/075051907; Daniel, Rolf; Dobrindt, Ulrich

    2016-01-01

    Escherichia coli is the leading causative agent of acute bovine mastitis. Here, we report the complete genome sequence of E. coli O70:H32 strain 1303, isolated from an acute case of bovine mastitis, and E. coli Ont:Hnt strain ECC-1470, isolated from a persistent infection.

  7. Preparation and evaluation of 68Ga-ECC as a PET renal imaging agent

    International Nuclear Information System (INIS)

    Mizaei, Alireza; Jaililan, Amir Reza; Mazidi, Mohammad; Aghanejad, Ayuob; Yousefnia, Hassan; Shabani, Gholamli; Ardaneh, Khosro; Geramifar, Patham; Beiki, Davood

    2015-01-01

    Development of a gallium-68-labeled renal tracer can be a good substitute for Tc-99m, a known SPECT tracer. In this study, effort was made to develop 68 Ga-ethylenecysteamine cysteine ( 68 Ga-ECC). Ga-ECC was prepared using generator-based 68 GaCl3 and ethylenecysteamine cysteine (ECC) at optimized conditions. Stability of the complex was checked in human serum followed by partition coefficient determination of the tracer. The biodistribution of the tracer in rats was studied using tissue counting and PET/CT imaging up to 120 min. Ga-ECC was prepared at optimized conditions in 15 min at 90 °C (radiochemical purity ≈97 ± 0.88 % ITLC, >99 % HPLC, specific activity: 210 ± 5 GBq/mM). 68 Ga-ECC was a water-soluble complex based on partition coefficient data (log P; −1.378) and was stable in the presence of human serum for 2 h at 37 °C. The biodistribution of the tracer demonstrated high kidney excretion of the tracer in 10–20 min. The SUV max ratios of the liver to left kidney were 0.38 and 0.39 for 30 and 90 min, respectively, indicating high kidney uptake. Initial biodistribution results showed significant kidney and urinary excretion of the tracer comparable to that of the homologous 99m Tc compound. The complex could be a possible PET kidney imaging agent with a fast imaging time

  8. Preparation and evaluation of {sup 68}Ga-ECC as a PET renal imaging agent

    Energy Technology Data Exchange (ETDEWEB)

    Mizaei, Alireza; Jaililan, Amir Reza; Mazidi, Mohammad; Aghanejad, Ayuob; Yousefnia, Hassan; Shabani, Gholamli; Ardaneh, Khosro [Radiation Application Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Geramifar, Patham; Beiki, Davood [Research Center for Nuclear Medicine, Shariati Hospital, Tehran University of Medical Sciences, Tehran (Iran, Islamic Republic of)

    2015-09-15

    Development of a gallium-68-labeled renal tracer can be a good substitute for Tc-99m, a known SPECT tracer. In this study, effort was made to develop {sup 68}Ga-ethylenecysteamine cysteine ({sup 68}Ga-ECC). Ga-ECC was prepared using generator-based {sup 68}GaCl3 and ethylenecysteamine cysteine (ECC) at optimized conditions. Stability of the complex was checked in human serum followed by partition coefficient determination of the tracer. The biodistribution of the tracer in rats was studied using tissue counting and PET/CT imaging up to 120 min. Ga-ECC was prepared at optimized conditions in 15 min at 90 °C (radiochemical purity ≈97 ± 0.88 % ITLC, >99 % HPLC, specific activity: 210 ± 5 GBq/mM). {sup 68}Ga-ECC was a water-soluble complex based on partition coefficient data (log P; −1.378) and was stable in the presence of human serum for 2 h at 37 °C. The biodistribution of the tracer demonstrated high kidney excretion of the tracer in 10–20 min. The SUV{sub max} ratios of the liver to left kidney were 0.38 and 0.39 for 30 and 90 min, respectively, indicating high kidney uptake. Initial biodistribution results showed significant kidney and urinary excretion of the tracer comparable to that of the homologous {sup 99m}Tc compound. The complex could be a possible PET kidney imaging agent with a fast imaging time.

  9. Cooling techniques

    International Nuclear Information System (INIS)

    Moeller, S.P.

    1994-01-01

    After an introduction to the general concepts of cooling of charged particle beams, some specific cooling methods are discussed, namely stochastic, electron and laser cooling. The treatment concentrates on the physical ideas of the cooling methods and only very crude derivations of cooling times are given. At the end three other proposed cooling schemes are briefly discussed. (orig.)

  10. Experiment data report for Semiscale Mod-1 Test S-05-4 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Collins, B.L.; Feldman, E.M.

    1977-03-01

    Recorded test data are presented for Test S-05-4 of the Semiscale Mod-1 alternate emergency core coolant injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-4 was conducted from initial conditions of 2266 psia and 543 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of each loop and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. The upper plenum coolant injection was scaled according to the heat stored in the metal mass of the upper plenum

  11. Reference core design Mark-I and -II of the experimental, multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    Shindo, Ryuiti; Hirano, Mitsumasa; Aruga, Takeo; Yasukawa, Sigeru

    1977-10-01

    Reactivity worth of the control rods and power distribution in the initial hot-clean core of reference core design Mark-I and -II have been studied. The need for burnable poison was confirmed, because of the limitations in number, diameter and reactivity worth of the control rods due to structures of pressure vessel and fuel element and to safety of the core. While the initial excess reactivity is reduced by use of the burnable poison, the recovery of core reactivity with burnup of the burnable poison requires a complicated withdrawal sequence of the control rods. The radial power gradient in the core is not large, due to orifice control of the coolant helium flow, effectiveness of the reflector in the small core and continuous distribution of burnup in the core by one-batch refuelling scheme. The local peaking factor in unit orifice regions, therefore, is the most important core design. Control of the axial power distribution is necessary to reduce the maximum fuel temperature and the exponential power distribution peaked toward the inlet of the core is most suitable. However, insertion of the control rods from top of the core disturbs the axial power distribution, so this effect must be considered in design of the withdrawal sequence of control rods. Nuclear properties of the core were revealed from results of the study for the initial hot-clean core. (auth.)

  12. Evaluation on Long-term Cooling of CANDU after Sump Blockage using MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Seon Oh; Cho, Yong Jin [KINS, Daejeon (Korea, Republic of); Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    There was a real incident that part of the fibrous insulation debris stripped by steam jet was transported to the pool and clogged the intake strainers of the drywell spray system, which revealed a weakness in the defense-in-depth concept which under other circumstances could have led to the ECCS failing to provide coolant to the core. Since the above Barseback-2 incident in 1992, lots of the international activities have been carried out to identify essential parameters and physical phenomena and to promote consensus on the technical issues, important for safety and possible paths for their resolution. In nuclear power plant under operation, if an unplanned reactor trip or a power reduction occurs, operators are required to maintain the reactor in a stable state according to emergency operating procedure (EOP) and to take diagnosis and appropriate mitigation actions if necessary. Subject to the EOP of Wolsong unit 1 (the first Korean PHWR NPP) under LOCA, intact or broken loops are diagnosed using the available plant information such as pressure and temperature of outlet headers. For the intact loop, effective long-term cooling is envisioned through the operation of shutdown cooling system as implemented in the EOP. In this work, the adequacy of long-term cooling during the recirculation phase of LOCA was evaluated under the postulated condition of the reduced flow path of the recirculation sump due to the inflow of substantial amount of debris released by the break flow with high energy. For the intact loop, although the incipience of boiling in the fuel channel was evaluated to occur, the effective long-term cooling can be achieved through the shutdown cooling system as guided in the EOP.

  13. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Vargas O, D.; Chavez M, C.

    2012-10-01

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

  14. Mitigation of hydrogen hazards in water cooled power reactors

    International Nuclear Information System (INIS)

    2001-02-01

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

  15. Rapid Endovascular Catheter Core Cooling Combined With Cold Saline as an Adjunct to Percutaneous Coronary Intervention for the Treatment of Acute Myocardial Infarction

    DEFF Research Database (Denmark)

    Erlinge, David; Götberg, Matthias; Lang, Irene

    2014-01-01

    : In a multicenter study, 120 patients with ST-segment elevation myocardial infarctions (rapid infusion of 600 to 2,000 ml cold saline and endovascular cooling or standard of care. Hypothermia was initiated...

  16. Cooling device for reactor container

    International Nuclear Information System (INIS)

    Arai, Kenji.

    1996-01-01

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

  17. Modifying the ECC-based grouping-proof RFID system to increase inpatient medication safety.

    Science.gov (United States)

    Ko, Wen-Tsai; Chiou, Shin-Yan; Lu, Erl-Huei; Chang, Henry Ker-Chang

    2014-09-01

    RFID technology is increasingly used in applications that require tracking, identification, and authentication. It attaches RFID-readable tags to objects for identification and execution of specific RFID-enabled applications. Recently, research has focused on the use of grouping-proofs for preserving privacy in RFID applications, wherein a proof of two or more tags must be simultaneously scanned. In 2010, a privacy-preserving grouping proof protocol for RFID based on ECC in public-key cryptosystem was proposed but was shown to be vulnerable to tracking attacks. A proposed enhancement protocol was also shown to have defects which prevented proper execution. In 2012, Lin et al. proposed a more efficient RFID ECC-based grouping proof protocol to promote inpatient medication safety. However, we found this protocol is also vulnerable to tracking and impersonation attacks. We then propose a secure privacy-preserving RFID grouping proof protocol for inpatient medication safety and demonstrate its resistance to such attacks.

  18. Scaling and design report of ECC performance test facility (SWAT) of SMART

    International Nuclear Information System (INIS)

    Cho, Seok; Ko, Yong Ju; Cho, Young Il; Kim, Jeong Tak; Choi, Nam Hyun; Shin Yong Chul; Park, Choon Kyong; Kwon, Tae Soon; Lee, Sung Jae

    2010-12-01

    SWAT (SMART ECC Water Asymmetric Two-phase choking test facility) was designed by 1/5 scaling ratio using the modified linear scaling method. The design characteristics of the SMART such that the elevation of RCP suction nozzles is the same with that of the ECC injection nozzles are maintained to reduce a distortion caused by the gravitational effect. Thermal hydraulic phenomena in a test facility designed by the modified linear scaling method can be simulated more accurately than those by the full-height and reduced area scaling method. The main part of the test section is SG-side upper down-comer. The boundary conditions are saturated steam and water flow condition and drain flow rate to control the collapsed water level in the down-comer. The test data of the SWAT can produce the well-defined boundary condition to validate the thermal hydraulic analysis code for the SMART

  19. Prosthetic Oral Rehabilitation of a Child With S-ECC: A Case Report with Histopathologic Analysis.

    Science.gov (United States)

    Tannure, P N; Moraes, G G; Borba, McU; Abrahão, A; Andrade, Mtrc; Fidalgo, Tks

    2015-01-01

    The aim of this case report is to describe the treatment planning of a young child with severe early childhood caries (S-ECC) as well as the prosthetic rehabilitation technique. A 3-year-old female child was referred to the pediatric dentistry clinic with the chief complaint of tooth pain, difficulty in eating and recurrent hospitalizations caused by dental infections. The mother reported intermittent episodes of fever and recurrent swelling of child's face. The girl presented angular cheilitis and was referred to a dietitian. The treatment plain consisted on a behavior changes in oral hygiene habits, exodontias of all primary teeth and oral rehabilitation with a prosthesis. The extracted teeth with periapical lesions were submitted to histopathologic analysis (hematoxilin and eosin staining) and revealed an inflammatory infiltrate. The aesthetic requirement of children with S-ECC has been a challenge to pediatric dentists. In the present case, the oral rehabilitation provided for the children better aesthetic, nutrition, phonation, and functional conditions.

  20. Scaling and design report of ECC performance test facility (SWAT) of SMART

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seok; Ko, Yong Ju; Cho, Young Il; Kim, Jeong Tak; Choi, Nam Hyun; Shin Yong Chul; Park, Choon Kyong; Kwon, Tae Soon; Lee, Sung Jae [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    SWAT (SMART ECC Water Asymmetric Two-phase choking test facility) was designed by 1/5 scaling ratio using the modified linear scaling method. The design characteristics of the SMART such that the elevation of RCP suction nozzles is the same with that of the ECC injection nozzles are maintained to reduce a distortion caused by the gravitational effect. Thermal hydraulic phenomena in a test facility designed by the modified linear scaling method can be simulated more accurately than those by the full-height and reduced area scaling method. The main part of the test section is SG-side upper down-comer. The boundary conditions are saturated steam and water flow condition and drain flow rate to control the collapsed water level in the down-comer. The test data of the SWAT can produce the well-defined boundary condition to validate the thermal hydraulic analysis code for the SMART

  1. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

    A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

  2. The E.C.C. programme in the field of radioactive wastes

    International Nuclear Information System (INIS)

    Orlowski, S.

    1976-01-01

    After a description of the aims of the E.C.C. in the field of radioactive wastes, the contents and the structure of the programme developed for this purpose as well as its time schedule and financing, are presented. The state of work progress and the first results of the actions taken, whose application is marked by an active cooperation of numerous national experts are quoted [fr

  3. Audit of ECCS Availability for CANDU Reactors with an extended O/H interval

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jong Soo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-10-15

    KINS conducts regulatory periodic inspections of the safety and performance of each nuclear installation during the planned outage every 20 months, pursuant to the Atomic Energy Act. For CANDU reactors, planned outage or overhaul (O/H) have been performed every 15 months. KHNP has been making efforts to extend the O/H intervals of CANDU reactors into 20 months since 2001. Low ECCS availability is one of the regulatory pending issues in the related licensing

  4. A Secure-Enhanced Data Aggregation Based on ECC in Wireless Sensor Networks

    Directory of Open Access Journals (Sweden)

    Qiang Zhou

    2014-04-01

    Full Text Available Data aggregation is an important technique for reducing the energy consumption of sensor nodes in wireless sensor networks (WSNs. However, compromised aggregators may forge false values as the aggregated results of their child nodes in order to conduct stealthy attacks or steal other nodes’ privacy. This paper proposes a Secure-Enhanced Data Aggregation based on Elliptic Curve Cryptography (SEDA-ECC. The design of SEDA-ECC is based on the principles of privacy homomorphic encryption (PH and divide-and-conquer. An aggregation tree disjoint method is first adopted to divide the tree into three subtrees of similar sizes, and a PH-based aggregation is performed in each subtree to generate an aggregated subtree result. Then the forged result can be identified by the base station (BS by comparing the aggregated count value. Finally, the aggregated result can be calculated by the BS according to the remaining results that have not been forged. Extensive analysis and simulations show that SEDA-ECC can achieve the highest security level on the aggregated result with appropriate energy consumption compared with other asymmetric schemes.

  5. Gonioscopic changes in conventional ECCE vs manual SICS: A comparative study

    Directory of Open Access Journals (Sweden)

    Surya Joseph, Sundararajan D, Rajvin Samuel Ponraj, Srinivasan M, Veluchamy

    2014-04-01

    Full Text Available The aim of the study is to observe and compare the Gonioscopic changes in the angle of the anterior chamber of the eye after surgeries namely; Conventional Extra capsular cataract extraction (ECCE with Posterior chamber Intraocular lens (PC IOL implantation, Manual Small incision Cataract Surgery with PCIOL implantation. The clinical study was undertaken after Institutional Ethical committee clearance, securing the inform consent, total number of 100 patients were enrolled in the study. 50 ECCE; 50 SICS consisting of 57 Males and 43 Females aged between 40 – 80yrs who were admitted and operated for Cataract at Meenakshi Medical college Hospital & Research institute. The following parameters are studied: Gonioscopic changes in the angle, namely the PAS formation in the quadrants, pigment dispersion in each of the methods. After this study, we arrive to a conclusion that complications in the angle of anterior chamber occur mostly in Conventional with insignificant change in manual SICS. So manual Small incision Cataract Surgery with PCIOL implantation is preferable over Conventional ECCE with PCIOL implantation.

  6. Fuel Cooling in Absence of Forced Flow at Shutdown Condition with PHTS Partially Drained

    Energy Technology Data Exchange (ETDEWEB)

    Parasca, L.; Pecheanu, D.L., E-mail: laurentiu.parasca@cne.ro, E-mail: doru.pecheanu@cne.ro [Cernavoda Nuclear Power Plant, Cernavoda (Romania)

    2014-09-15

    During the plant outage for maintenance on primary side (e.g. for the main Heat Transport System pumps maintenance, the Steam Generators inspection), there are situations which require the primary heat transport system (HTS) drainage to a certain level for opening the circuit. The primary fuel heat sink for this configuration is provided by the shutdown cooling system (SDCS). In case of losing the forced cooling (e.g. due to the loss of SDCS, design basis earthquake-DBE), flow conditions in the reactor core may become stagnant. Inside the fuel channels, natural circulation phenomena known as Intermittent Buoyancy Induced Flow (IBIF) will initiate, providing an alternate heat sink mechanism for the fuel. However, this heat sink is effective only for a limited period of time (recall time). The recall time is defined as the elapsed time until the water temperature in the HTS headers exceeds a certain limit. Until then, compensatory measures need to be taken (e.g. by re-establishing the forced flow or initiate Emergency Core Cooling system injection) to preclude fuel failures. The present paper briefly presents the results of an analysis performed to demonstrate that fuel temperature remains within acceptable limits during IBIF transient. One of the objectives of this analysis was to determine the earliest moment since the reactor shut down when maintenance activities on the HTS can be started such that IBIF is effective in case of losing the forced circulation. The resulting peak fuel sheath and pressure tube temperatures due to fuel heat up shall be within the acceptable limits to preclude fuel defect or fuel channel defects.Thermalhydraulic circuit conditions were obtained using a CATHENA model for the primary side of HTS (drained to a certain level), an ECC system model and a system model for SDCS. A single channel model was developed in GOTHIC code for the fuel assessment analysis. (author)

  7. Quorum sensing in the plant pathogen Erwinia carotovora subsp. carotovora: the role of expR(Ecc).

    Science.gov (United States)

    Andersson, R A; Eriksson, A R; Heikinheimo, R; Mäe, A; Pirhonen, M; Kõiv, V; Hyytiäinen, H; Tuikkala, A; Palva, E T

    2000-04-01

    The production of the main virulence determinants of the plant pathogen Erwinia carotovora subsp. carotovora, the extracellular cell wall-degrading enzymes, is partly controlled by the diffusible signal molecule N-(3-oxohexanoyl)-L-homoserine lactone (OHHL). OHHL is synthesized by the product of the expI/carI gene. Linked to expI we found a gene encoding a putative transcriptional regulator of the LuxR-family. This gene, expR(Ecc), is transcribed convergently to the expI gene and the two open reading frames are partially overlapping. The ExpR(Ecc) protein showed extensive amino acid sequence similarity to the repressor EsaR from Pantoea stewartii subsp. stewartii (formerly Erwinia stewartii subsp. stewartii) and to the ExpR(Ech) protein of Erwinia chrysanthemi. Inactivation of the E. carotovora subsp. carotovora expR(Ecc) gene caused no decrease in virulence or production of virulence determinants in vitro. In contrast, there was a slight increase in the maceration capacity of the mutant strain. The effects of ExpR(Ecc) were probably mediated by changes in OHHL levels. Inactivation of expR(Ecc) resulted in increased OHHL levels during early logarithmic growth. In addition, overexpression of expR(Ecc) caused a clear decrease in the production of virulence determinants and part of this effect was likely to be caused by OHHL binding to ExpR(Ecc). ExpR(Ecc) did not appear to exhibit transcriptional regulation of expI, but the effect on OHHL was apparently due to other mechanisms.

  8. Intricate and Cell Type-Specific Populations of Endogenous Circular DNA (eccDNA) in Caenorhabditis elegans and Homo sapiens.

    Science.gov (United States)

    Shoura, Massa J; Gabdank, Idan; Hansen, Loren; Merker, Jason; Gotlib, Jason; Levene, Stephen D; Fire, Andrew Z

    2017-10-05

    Investigations aimed at defining the 3D configuration of eukaryotic chromosomes have consistently encountered an endogenous population of chromosome-derived circular genomic DNA, referred to as extrachromosomal circular DNA (eccDNA). While the production, distribution, and activities of eccDNAs remain understudied, eccDNA formation from specific regions of the linear genome has profound consequences on the regulatory and coding capabilities for these regions. Here, we define eccDNA distributions in Caenorhabditis elegans and in three human cell types, utilizing a set of DNA topology-dependent approaches for enrichment and characterization. The use of parallel biophysical, enzymatic, and informatic approaches provides a comprehensive profiling of eccDNA robust to isolation and analysis methodology. Results in human and nematode systems provide quantitative analysis of the eccDNA loci at both unique and repetitive regions. Our studies converge on and support a consistent picture, in which endogenous genomic DNA circles are present in normal physiological states, and in which the circles come from both coding and noncoding genomic regions. Prominent among the coding regions generating DNA circles are several genes known to produce a diversity of protein isoforms, with mucin proteins and titin as specific examples. Copyright © 2017 Shoura et al.

  9. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Bolt, S.E.

    1977-01-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and four thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. In the first experiment, initiation was not expected and did not occur, although there was a small amount of subcritical crack growth. In the second experiment, initiation of a semicircular flaw took place as expected; the final length along the surface was about four times the initial length, but there was no radial growth. The third and fourth experiments were similar, and the long axial flaw initiated in good agreement with predictions

  10. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    International Nuclear Information System (INIS)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm 2 , 1000 0 C cladding temperature, and (2) 40 h at 40 W/cm 2 , 1200 0 C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370 0 C

  11. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  12. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

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

  13. Evaluation report on CCTF Core-II reflood test second shakedown test, C2-SH2 (Run 54)

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Sugimoto, Jun; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio

    1985-03-01

    A low power test (the initial averaged linear power density = 1.18 kW/m) and the base case test (1.4 kW/m) were performed with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute, in order to study the effect of the power on the reflood phenomena. The former linear power density corresponds nearly to the scaled linear power density based on the current safety evaluation criterio. During the early period of the reflood ( 200s) the heat transfer coefficient became higher and resultantly the quench front advanced faster in the low power test. The core flooding rate was nearly identical between both tests, independently of the different power. The insensitiveness of the power to the core flooding rate was also observed in FLECHT-SET performed in the USA. A significatn large differential pressure oscillation at ECC ports was experienced in the low power test, and it may be important for the long term core cooling although it has not been taken note on the previous studies. (author)

  14. 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.

  15. 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.

  16. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (Dpower is 300 MWth. First calculation, we calculate survey parameter for UN-PuN fuel with fissile contain from Plutonium waste LWR for each geometry. The minimum power density is around 72 Watt/cc, and maximum power density 114 Watt/cc. After we calculate with various geometry core, when we use the balance geometry, the k-eff value flattest and more stable than the others.

  17. Cool Science Explains a Warming World: Using Ice Core Science to Bridge the Gap Between Researchers and the K-12 Classroom

    Science.gov (United States)

    Huffman, L. T.

    2017-12-01

    Changing ice has urgent implications for people around the world. The Ice Drilling Program Office (IDPO) provides scientific leadership and oversight of ice coring and drilling activities funded by the US National Science Foundation and also has goals to enhance education and communication of current research information. In a time when misinformation is rampant and climate change science is suspect, it is essential that students receive accurate scientific information and engage in learning activities that model complex ideas through engaging and age appropriate ways, while also learning to validate and recognize reliable sources. The IDPO Education and Outreach (EO) office works to create resources, activities and professional development that bridge the gap between ice core science research and educators and their students. Ice core science is on the cutting edge of new discoveries about climate change and understanding better the past to predict the future. Hands-on inquiry activities based on ice core data allow teachers to lead their students to new discoveries about climate secrets hidden deep in the ice. Capitalizing on the inherent interest in the extremes of the Polar Regions, IDPO materials engage students in activities aligned with NGSS standards. Ice drilling technologies make an ideal platform for intertwining engineering concepts and practices with science research to meet the SEP (Science and Engineering Practices) in the NGSS. This session will highlight how the IDPO EO office has built a community of ice core scientists willing to take part in education and outreach projects and events and share some of the resources available to K-12 educators. We will highlight some of the successes and lessons learned as we continually evolve our work toward more effective science education and communication highlighting ice core and climate change science.

  18. Prevention and investigations of core degradation in case of beyond design accidents of the 2400 MWTH gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Bertrand, F.; Gatin, V.; Bentivoglio, F.; Gueneau, C.

    2011-01-01

    The present paper deals with studies carried out to assess the ability of the core of the Gas Fast Reactor (GFR) to withstand beyond design accidents. The work presented here is aimed at simulating the behaviour of this core by using analytical models whose input parameters are calculated with the CATHARE2 code. Among possible severe accident initiators, the Unprotected Loss Of Coolant Accident (ULOCA of 3 Inches diameter) is investigated in detail in the paper with CATHARE2. Additionally, a simplified pessimistic assessment of the effect of a postulated power excursion that could result from the failure of prevention provisions is presented. (author)

  19. On the energy shift of the ECC cusp. Does the shift really exist?

    International Nuclear Information System (INIS)

    Barrachina, R.O.; Sarkadi, L.

    2004-01-01

    Full text: The cusplike 'electron capture to the continuum' (ECC) peak appearing in the spectrum of the forward emitted electrons in ion-atom collisions is generally thought to be a divergence. The peak is centered at electron velocity that matches that of the projectile. Recently Illescas et al. [1] claimed, however, that 'the cusp is not a divergence smoothed by the experiment, and is slightly shifted from the impact-velocity value'. In a subsequent work Shah et al. [2] measured the ECC cusp for collisions of 10- and 20- keV protons with H2 and He, and found that the peak was indeed shifted to a velocity below its expected position. Their classical trajectory Monte Carlo (CTMC) calculations reproduced the observation well. In the present work we demonstrate by our CTMC calculations (made for the case of 20-keV protons on He) that the intensity, width and position of the ECC cusp do depend on the experimental conditions by which the electrons are detected (first of all, the acceptance angle of the electron spectrometer). We also present experimental data for this collision system. While the latter data are well reproduced by our CTMC calculations, neither the theoretical results, nor the experimental data of the present work support of the findings of Shah et al. [2]. At the same time, analyzing the problem in the framework of the general final-state interaction theory of cusp formation [3], we concluded that 'rigorously' speaking Shah et al. [2] are right when they state that the ECC cusp is not a divergence. The divergence is smeared out by the scattering of the projectile. However, this effect is negligibly small at the lowest energy that was considered in their paper, 10 keV. At this impact energy the projectile motion is smeared within an angular region Δθ th ∼ 0.06 deg, which is much smaller than the acceptance angle of the experiment, Δθ exp ∼ 1.5 deg

  20. Assessment of compliance with regulatory requirements for a best estimate methodology for evaluation of ECCS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Un Chul; Jang, Jin Wook; Lim, Ho Gon; Jeong, Ik [Seoul National Univ., Seoul (Korea, Republic of); Sim, Suk Ku [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-03-15

    Best estimate methodology for evaluation of ECCS proposed by KEPCO(KREM) os using thermal-hydraulic best-estimate code and the topical report for the methodology is described that it meets the regulatory requirement of USNRC regulatory guide. In this research the assessment of compliance with regulatory guide. In this research the assessment of compliance with regulatory requirements for the methodology is performed. The state of licensing procedure of other countries and best-estimate evaluation methodologies of Europe is also investigated, The applicability of models and propriety of procedure of uncertainty analysis of KREM are appraised and compliance with USNRC regulatory guide is assessed.

  1. A methodology for the estimation of release of fission products during LOCA with loss of ECCS

    International Nuclear Information System (INIS)

    Lele, H.G.; Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K.; Venkat Raj, V.

    2002-01-01

    A Loss of Coolant Accident (LOCA) in a nuclear reactor along with the failure of the Emergency Core Cooling System can cause sustained voiding of the core. In such a situation the core experiences very low flow which leads to poor heat removal from the reactor core. The heat to be removed from the core includes stored heat, heat generated due to metal water reaction at high temperatures, decay heat etc. The poor heat removal leads to heating of the fuel pins to high temperatures. The heating of fuel pins is further enhanced due to metal-water reaction at high temperatures. These high temperatures of the fuel pins can lead to fission product release, which is transported into the Primary Heat Transport (PHT) system and can enter the containment through the break. Analysis is involved due to the complexity of the system and the phenomena to be simulated. In this paper a multistage analysis methodology is presented that involves the development and application of a number of computer programs to model the various phenomena involved. The computer code PHTACT computes the activity release from the fuel as a function of fuel temperatures and cladding oxidation, its distribution into the PHT system and release into the containment. Computation of thermal hydraulic parameters during LOCA is done using the thermal hydraulic analysis code RELAP5. The detailed simulation of fuel pin temperatures is done using computer code HT/MOD4. The convective boundary conditions required for the code are obtained from RELAP5. Creep deformation is considered in the computation of dimensional changes of the coolant channel and estimation of flow blockage due to clad ballooning. The progression of various reaction layers due to high temperature reaction between fuel and clad and clad and steam is also computed, which affects the structural strength of the clad. Different approaches are possible and analysis can be carried out in different phases depending upon the complexities to be

  2. Cooling rates and the depth of detachment faulting at oceanic core complexes: Evidence from zircon Pb/U and (U-Th)/He ages

    Science.gov (United States)

    Grimes, Craig B.; Cheadle, Michael J.; John, Barbara E.; Reiners, P.W.; Wooden, J.L.

    2011-01-01

    Oceanic detachment faulting represents a distinct mode of seafloor spreading at slow spreading mid-ocean ridges, but many questions persist about the thermal evolution and depth of faulting. We present new Pb/U and (U-Th)/He zircon ages and combine them with magnetic anomaly ages to define the cooling histories of gabbroic crust exposed by oceanic detachment faults at three sites along the Mid-Atlantic Ridge (Ocean Drilling Program (ODP) holes 1270D and 1275D near the 15??20???N Transform, and Atlantis Massif at 30??N). Closure temperatures for the Pb/U (???800??C-850??C) and (U-Th)/He (???210??C) isotopic systems in zircon bracket acquisition of magnetic remanence, collectively providing a temperature-time history during faulting. Results indicate cooling to ???200??C in 0.3-0.5 Myr after zircon crystallization, recording time-averaged cooling rates of ???1000??C- 2000??C/Myr. Assuming the footwalls were denuded along single continuous faults, differences in Pb/U and (U-Th)/He zircon ages together with independently determined slip rates allow the distance between the ???850??C and ???200??C isotherms along the fault plane to be estimated. Calculated distances are 8.4 ?? 4.2 km and 5.0 2.1 km from holes 1275D and 1270D and 8.4 ?? 1.4 km at Atlantis Massif. Estimating an initial subsurface fault dip of 50 and a depth of 1.5 km to the 200??C isotherm leads to the prediction that the ???850??C isotherm lies ???5-7 km below seafloor at the time of faulting. These depth estimates for active fault systems are consistent with depths of microseismicity observed beneath the hypothesized detachment fault at the TAG hydrothermal field and high-temperature fault rocks recovered from many oceanic detachment faults. Copyright 2011 by the American Geophysical Union.

  3. Cooling towers

    International Nuclear Information System (INIS)

    Boernke, F.

    1975-01-01

    The need for the use of cooling systems in power plant engineering is dealt with from the point of view of a non-polluting form of energy production. The various cooling system concepts up to the modern natural-draught cooling towers are illustrated by examples. (TK/AK) [de

  4. A provably-secure ECC-based authentication scheme for wireless sensor networks.

    Science.gov (United States)

    Nam, Junghyun; Kim, Moonseong; Paik, Juryon; Lee, Youngsook; Won, Dongho

    2014-11-06

    A smart-card-based user authentication scheme for wireless sensor networks (in short, a SUA-WSN scheme) is designed to restrict access to the sensor data only to users who are in possession of both a smart card and the corresponding password. While a significant number of SUA-WSN schemes have been suggested in recent years, their intended security properties lack formal definitions and proofs in a widely-accepted model. One consequence is that SUA-WSN schemes insecure against various attacks have proliferated. In this paper, we devise a security model for the analysis of SUA-WSN schemes by extending the widely-accepted model of Bellare, Pointcheval and Rogaway (2000). Our model provides formal definitions of authenticated key exchange and user anonymity while capturing side-channel attacks, as well as other common attacks. We also propose a new SUA-WSN scheme based on elliptic curve cryptography (ECC), and prove its security properties in our extended model. To the best of our knowledge, our proposed scheme is the first SUA-WSN scheme that provably achieves both authenticated key exchange and user anonymity. Our scheme is also computationally competitive with other ECC-based (non-provably secure) schemes.

  5. Experimental results of the SMART ECC injection performance with reduced scale of test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Il; Cho, Seok; Ko, Yung Joo; Shin, Yong Cheol; Kwon, Tae Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    SMART pressurized water reactor type is different from the existing integral NSSS commercial pressurized water reactor system which is equipped with the main features. In addition, RCS piping is removed and the feature of the SBLOCA is a major design break accident. SWAT (SMART ECC Water Asymmetric Two-phase choking test facility) test facility is to simulate the 2 inch SBLOCA of the SMART using with reduced scale. The Test was performed to produce experimental data for the validation of the TASS/SMR-S thermal hydraulic analysis code, and to investigate the related thermal hydraulic phenomena in the down-comer region during the 2 inch SBLOCA of the safety inject line. The particular phenomena for the observation are ECC bypass and multi-dimensional flow characteristics to verify the effectiveness and performance of the safety injection system. In this paper, the corresponding steady state test conditions, including initial and boundary conditions along with major measuring parameters, and related experimental results were described

  6. Assessment of human effective absorbed dose of 67 Ga-ECC based on biodistribution rat data.

    Science.gov (United States)

    Shanehsazzadeh, Saeed; Yousefnia, Hassan; Lahooti, Afsaneh; Zolghadri, Samaneh; Jalilian, Amir Reza; Afarideh, Hossien

    2015-02-01

    In a diagnostic context, determination of absorbed dose is required before the introduction of a new radiopharmaceutical to the market to obtain marketing authorization from the relevant agencies. In this work, the absorbed dose of [67 Ga]-ethylenecysteamine cysteine [(67 Ga)ECC] to human organs was determined by using distribution data for rats. For biodistribution data, the animals were sacrificed by CO2 asphyxiation at selected times after injection (0.5, 2 and 48 h, n = 3 for each time interval), then the tissue (blood, heart, lung, brain, intestine, feces, skin, stomach, kidneys, liver, muscle and bone) were removed. The absorbed dose was determined by Medical Internal Radiation Dose (MIRD) method after calculating cumulated activities in each organ. Our prediction shows that a 185-MBq injection of (67)Ga-ECC into the humans might result in an estimated absorbed dose of 0.029 mGy in the whole body. The highest absorbed doses are observed in the spleen and liver with 33.766 and 16.847 mGy, respectively. The results show that this radiopharmaceutical can be a good SPECT tracer since it can be produced easily and also the absorbed dose in each organ is less than permitted absorbed dose.

  7. A Provably-Secure ECC-Based Authentication Scheme for Wireless Sensor Networks

    Science.gov (United States)

    Nam, Junghyun; Kim, Moonseong; Paik, Juryon; Lee, Youngsook; Won, Dongho

    2014-01-01

    A smart-card-based user authentication scheme for wireless sensor networks (in short, a SUA-WSN scheme) is designed to restrict access to the sensor data only to users who are in possession of both a smart card and the corresponding password. While a significant number of SUA-WSN schemes have been suggested in recent years, their intended security properties lack formal definitions and proofs in a widely-accepted model. One consequence is that SUA-WSN schemes insecure against various attacks have proliferated. In this paper, we devise a security model for the analysis of SUA-WSN schemes by extending the widely-accepted model of Bellare, Pointcheval and Rogaway (2000). Our model provides formal definitions of authenticated key exchange and user anonymity while capturing side-channel attacks, as well as other common attacks. We also propose a new SUA-WSN scheme based on elliptic curve cryptography (ECC), and prove its security properties in our extended model. To the best of our knowledge, our proposed scheme is the first SUA-WSN scheme that provably achieves both authenticated key exchange and user anonymity. Our scheme is also computationally competitive with other ECC-based (non-provably secure) schemes. PMID:25384009

  8. A Provably-Secure ECC-Based Authentication Scheme for Wireless Sensor Networks

    Directory of Open Access Journals (Sweden)

    Junghyun Nam

    2014-11-01

    Full Text Available A smart-card-based user authentication scheme for wireless sensor networks (in short, a SUA-WSN scheme is designed to restrict access to the sensor data only to users who are in possession of both a smart card and the corresponding password. While a significant number of SUA-WSN schemes have been suggested in recent years, their intended security properties lack formal definitions and proofs in a widely-accepted model. One consequence is that SUA-WSN schemes insecure against various attacks have proliferated. In this paper, we devise a security model for the analysis of SUA-WSN schemes by extending the widely-accepted model of Bellare, Pointcheval and Rogaway (2000. Our model provides formal definitions of authenticated key exchange and user anonymity while capturing side-channel attacks, as well as other common attacks. We also propose a new SUA-WSN scheme based on elliptic curve cryptography (ECC, and prove its security properties in our extended model. To the best of our knowledge, our proposed scheme is the first SUA-WSN scheme that provably achieves both authenticated key exchange and user anonymity. Our scheme is also computationally competitive with other ECC-based (non-provably secure schemes.

  9. Influence of time dependent longitudinal magnetic fields on the cooling process, exchange bias and magnetization reversal mechanism in FM core/AFM shell nanoparticles: a Monte Carlo study.

    Science.gov (United States)

    Yüksel, Yusuf; Akıncı, Ümit

    2016-12-07

    Using Monte Carlo simulations, we have investigated the dynamic phase transition properties of magnetic nanoparticles with ferromagnetic core coated by an antiferromagnetic shell structure. Effects of field amplitude and frequency on the thermal dependence of magnetizations, magnetization reversal mechanisms during hysteresis cycles, as well as on the exchange bias and coercive fields have been examined, and the feasibility of applying dynamic magnetic fields on the particle have been discussed for technological and biomedical purposes.

  10. Factors predisposing to early childhood caries (ECC) in children of pre-school age in the city of Zagreb, Croatia.

    Science.gov (United States)

    Lulić-Dukić, O; Jurić, H; Dukić, W; Glavina, D

    2001-06-01

    The aim of this study was to investigate factors predisposing to early childhood caries (ECC) in pre-school children in the city of Zagreb, Croatia. The investigation was carried out on the sample of 145 children (77 boys and 68 girls) aged between 2 and 5 years, including clinical examination of dental status and survey on the habits among the parents. The overall prevalence of ECC was 30%: in girls it was 25%, and in boys 48%. The study on the risk factors was designed as a classic case-control study. The mean value of dmfs index among the cases amounted to 8.6, in comparison to 5.2 in the control group (pteeth-brushing habit after first 24 months did (p importance of early introduction of teeth-brushing and giving up the nightly consumption of sweet beverages in prevention of ECC.

  11. Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

    International Nuclear Information System (INIS)

    Murao, Y.; Iguchi, T.; Sugimoto, J.

    1988-09-01

    The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

  12. 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

  13. Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores; Homogeneisation de modeles de transferts thermiques et radiatifs: application au coeur des reacteurs a caloporteur gaz

    Energy Technology Data Exchange (ETDEWEB)

    El Ganaoui, K

    2006-09-15

    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)

  14. Thermodynamic properties of helium in the range from 20 to 15000C and 1 to 100 bar. Reactor core design of high-temperature gas-cooled reactors. Pt. 1

    International Nuclear Information System (INIS)

    Kipke, H.E.; Stoehr, A.; Banerjea, A.; Hammeke, K.; Huepping, N.

    1978-12-01

    The following report presents in tabular form the safety standard of the nuclear safety standard commission (KTA) on reactor core design of high-temperature gas-cooled reactors. Part 1: Calculation of thermodynamic properties of helium The basis of the present work is the data and formulae given by H. Petersen for the calculation of density, specific heat, thermal conductivity and dynamic viscosity of helium together with the formula for their standard deviations in the range of temperature and pressure stated above. The relations for specific enthalpy and specific entropy have been derived from density and specific heat, whereby specific heat is assumed constant over the given range of temperature and pressure. The latter section of this report contains tables of thermodynamic properties of helium calculated from the equations stated earlier in this paper. (orig.) [de

  15. Assessment of some interfacial shear correlations in a model of ECC bypass flow in PWR reactor downcomer

    International Nuclear Information System (INIS)

    Popov, N.K.; Rohatgi, U.S.

    1987-01-01

    The bypass/refill process in the PWR reactor downcomer, following a large rupture of a cold leg coolant supply pipe, is a complicated thermo-hydraulic two-phase flow phenomenon. Mathematical modeling of such phenomena is always accompanied with a difficult task of selection of suitable constitutive correlations. In a typically hydrodynamic phenomenon, like ECC refill process of the reactor lower plenum is considered, the phasic interfacial friction is the most influential constitutive correlation. Therefore, assessment of the well-known widely-used interfacial friction constitutive correlations in the model of ECC bypass/refill process, is the subject of this paper

  16. Mechanical interaction of Engineered Cementitious Composite (ECC) reinforced with Fiber Reinforced Polymer (FRP) rebar in tensile loading

    DEFF Research Database (Denmark)

    Lárusson, Lárus Helgi; Fischer, Gregor; Jönsson, Jeppe

    2010-01-01

    This paper introduces a preliminary study of the composite interaction of Engineered Cementitious Composite (ECC), reinforced with Glass Fiber Reinforced Polymer (GFRP) rebar. The main topic of this paper will focus on the interaction of the two materials (ECC and GFRP) during axial loading......, particularly in post cracking phase of the concrete matrix. The experimental program carried out in this study examined composite behavior under monotonic and cyclic loading of the specimens in the elastic and inelastic deformation phases. The stiffness development of the composite during loading was evaluated...

  17. Factors Predisposing to Early Childhood Caries (ECC) in Children of Pre-School Age in the City of Zagreb, Croatia

    OpenAIRE

    Lulić-Dukić, O.; Jurić, H.; Dukić, W.; Glavina, D.

    2001-01-01

    The aim of this study was to investigate factors predisposing to early childhood caries (ECC) in pre-school children in the city of Zagreb, Croatia. The investigation was carried out on the sample of 145 children (77 boys and 68 girls) aged between 2 and 5 years, including clinical examination of dental status and survey on the habits among the parents. The overall prevalence of ECC was 30%: in girls it was 25%, and in boys 48%. The study on the risk factors was designed as a c...

  18. Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP; Simulacion del sistema de aspersion del nucleo alta presion (HPCS) para un reactor de agua en ebullicion (BWR) basado en RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas O, D.; Chavez M, C., E-mail: danmirnyi@gmail.com [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2012-10-15

    A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant s